DCL-03-183, License Amendment Request 03-18, Revision to Technical Specifications 5.5.9, Steam Generator (SG Tube Surveillance Program, and 5.6.10, SG Tube Inspection Report, for 4-Volt Alternate Repair Criteria for SG Tube Repair

License Amendment Request 03-18, Revision to Technical Specifications 5.5.9, Steam Generator (SG Tube Surveillance Program, and 5.6.10, SG Tube Inspection Report, for 4-Volt Alternate Repair Criteria for SG Tube Repair

ML040120619
Person / Time
Site:	Diablo Canyon
Issue date:	01/07/2004
From:	Oatley D Pacific Gas & Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-03-183
Download: ML040120619 (80)

Text

Pacific Gas and Electic Company David H. Oatley Diablo Canyon Power Plant Vice President and PO. Box 56 General Manager Avila Beach, CA 93424 January 7, 2004 805.545.4350 Fax: 805.545.4234 PG&E Letter DCL-03-183 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units I and 2 License Amendment Request 03-18 Revision to Technical Specifications 5.5.9. "Steam Generator (SG) Tube Surveillance Program," and 5.6.10, "Steam Generator (SG) Tube Inspection Report,"

for 4-volt Alternate Repair Criteria for Steam Generator Tube Repair In accordance with 10 CFR 50.90, enclosed is an application for amendment to Facility Operating License Nos. DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP) respectively. The enclosed license amendment request (LAR) proposes to revise the DCPP Technical Specifications (TS) to allow application of a 4-volt alternate repair criteria (ARC) at intersections of steam generator (SG) tube hot-legs with the 4 lowest SG tube support plates (TSPs). The 4-volt ARC will only apply to Model 51 SG tubes experiencing outside diameter stress corrosion cracking (ODSCC) at the intersections at the 4 lowest hot-leg TSPs.

The 4-volt ARC is based upon SG tube expansion at TSP intersections. With tube expansion at TSP intersections, TSP displacements during a postulated main steam line break event are limited to negligible levels. The number of SG tube expansions required is based on the maximum thermal hydraulic loads on the SG TSPs during a postulated main steam line break event.

This LAR proposes changes to TS 5.5.9, "Steam Generator (SG) Tube Surveillance Program," pages 5.0-11, 5.0-13a, 5.0-14, and 5.0-15, and TS 5.6.10, "Steam Generator (SG) Tube Inspection Report," page 5.0-29. A change to TS page 5.0-14 has also been proposed in PG&E Letter DCL-03-174, "License Amendment Request 03-16, Revised Wedge Region Exclusion Zones for Steam Generator Tube Alternate Repair Criteria,' dated December 19, 2003. A change to TS pages 5.0-15 and 5.0-29 have previously been proposed in PG&E Letter DCL-03-132, "License Amendment Request 03-15, Steam Generator Tube Repair Using Leak Limiting Alloy 800 Sleeves and Revision to Technical Specification Table 5.5.9-2, 'Steam Generator (SG) Tube Inspection,'" dated October 22, 2003.

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Document Control Desk PG&E Letter DCL-03-183 January 7, 2004 Page 2 In addition, the proposed change includes the application of leak-before-break (LBB) to the main steam line piping inside containment in order to exclude the dynamic effects of a main steam line break upstream of the flow restrictor from consideration for determining the loads on the SG TSPs following a steam line break event. As required by General Design Criteria (GDC) 4 of Appendix A to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50), staff review and approval of the LBB analysis is required before a licensee can exclude the dynamic effects from the design basis. LBB will only be applied to limit the thermal hydraulic analysis to small main steam line breaks to determine the loads on the SG TSPs to support the 4-volt ARC. The proposed LBB to the main steam lines will only be applied for this purpose until the DCPP SGs are replaced.

The DCPP SGs have experienced tube degradation related to axial ODSCC.

Currently, a SG tube with axial ODSCC degradation confined within the thickness of the SG TSP with a bobbin coil probe voltage greater than 2 volts is removed from service by plugging the tube on both ends. However, the installation of plugs in a SG tube eliminates the heat transfer surface associated with the tube. In addition, plug installation leads to the reduction in the primary coolant flow available for core cooling. Therefore, the installation of plugs into SG tubes is currently limited to a maximum of 15 percent of the tubes in any SG. The use of a 4-volt axial ODSCC ARC in the SG tube hot-leg region of the 4 lowest TSPs would allow SG tubes with bobbin indications up to 4 volts to remain in service and thus minimize the reduction in SG tube heat transfer surface area and primary coolant flow due to tube plugging.

The proposed 4-volt ARC in this LAR is similar to the 3-volt ARC approved by the NRC for South Texas Project Nuclear Operating Company's South Texas Project Unit 2 in Amendment 114 to Facility Operating License No. NPF-80, "South Texas Project (STP) Unit 2 - Issuance of Amendment Revising the Technical Specifications to Implement 3-volt Alternate Repair Criteria for Steam Generator Tube Repair (TAC No. MA8271)," dated March 8, 2001. The NRC has also previously approved a similar 3-volt ARC for Cycle 6 of Excelon's Braidwood Nuclear Power Station Unit 1 in License Amendment No. 69 to Facility Operating License No. NPF-72 and for Cycle 8 of the Byron Nuclear Power Station Unit 1 in License Amendment No. 77 to Facility Operating License No. NPF-37, "Issuance of Amendments (TAC Nos.

M96498, M96499, M96500 and M96501)," dated November 9, 1995, and for Cycle 7 of the Braidwood 1 Nuclear Power Station Unit 1 in License Amendment No. 82 to Facility Operating License No. NPF-72, "Issuance of Amendments (TAC Nos.

M91671, M91672, M91673 and M91674)," dated May 14,1997. The NRC has previously approved the application of LBB to main steam line piping inside A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

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Document Control Desk PG&E Letter DCL-03-183 January 7, 2004 Page 3 containment for the AP600 Standard Design in NUREG-1512, "Final Safety Evaluation Report Related to the Certification of the AP600 Standard Design, Docket No.52-003," dated August 1998, and for the System 80+ Design in NUREG-1462, Volume 1, 'Final Safety Evaluation Report Related to the Certification of the System 80+ Design, Docket No.52-002," dated August 1994.

Except for the magnitude of the bobbin coil voltage repair limit and number of cycles for which the repair limit is applicable, the proposed 4-volt repair limit is similar to the 3-volt repair limits approved for South Texas Project Unit 2, Byron Unit 1, and Braidwood Unit 1. The magnitude of the proposed 4-volt ARC is a 2 volt increase over the currently licensed GL 95-05 2-volt repair limit for 0.875 inch outside diameter Alloy 600 SG tubes. Similarly, the 3-volt ARC approved for South Texas Project Unit 2, Byron Unit 1, and Braidwood Unit I was a 2 volt increase over the previously licensed GL 95-05 1-volt repair limit for 0.75 inch outside diameter Alloy 600 SG tubes. contains a description of the proposed change, the supporting technical analyses, and the no significant hazards consideration determination. Enclosures 2 and 3 contain marked-up and retyped (clean) TS pages, respectively.

The technical basis for the 4-volt ARC is described in Westinghouse Electric LLC WCAP-16170-NP, Revision 0, "Diablo Canyon SG Alternate Repair Criteria Based on Limited Tube Support Plate Displacement," dated November 2003 (nonproprietary), contained in Enclosure 4 of this letter and WCAP-16170-P, Revision 0, "Diablo Canyon SG Alternate Repair Criteria Based on Limited Tube Support Plate Displacement," dated November 2003 (proprietary), contained in of this letter.

WCAP-1 6170-P, Revision 0, contains information proprietary to Westinghouse Electric Company LLC ("Westinghouse"). Accordingly, Enclosure 5 includes a Westinghouse authorization letter, CAW-03-1741, an accompanying affidavit, a Proprietary Information Notice, and a Copyright Notice. The affidavit is signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the Westinghouse proprietary information contained in WCAP-16170-P, Revision 0, may be withheld from public disclosure by the Commission, and it addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.790 of the Commission's regulations. PG&E requests that the Westinghouse proprietary information be withheld from public disclosure in accordance with 10 CFR 2.790.

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Document Control Desk PG&E Letter DCL-03-183 January 7, 2004 Page 4 Correspondence with respect to the copyright or proprietary aspects of the application for withholding related to the Westinghouse proprietary information or the Westinghouse affidavit provided in Enclosure 5 should reference Westinghouse Letter CAW-03-1741 and be addressed to J. S. Galembush, Acting Manager of Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania, 15230-0355.

A summary report of the LBB analysis for the main steam line, which supports exclusion of large main steam line breaks inside containment, is contained in to this letter.

PG&E has determined that this LAR does not involve a significant hazard consideration as determined per 10 CFR 50.92. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

The change in this LAR is not required to address an immediate safety concern.

However, PG&E requests approval of this LAR no later than October 1, 2004, to support the upcoming twelfth DCPP Unit 2 refueling outage (2R12) currently scheduled for October 2004. Approval of this change will reduce the number of SG tubes which need to be removed from service by plugging during 2R12 and will also allow certain SG tubes which are currently plugged to be returned to service by removing the plugs. PG&E requests the LAR be made effective upon NRC issuance, to be implemented within 60 days from the date of issuance.

If you have any questions or require additional information, please contact Stan Ketelsen at 805-545-4720.

Sincerely, David H. Oatley Vice President and General Manager - Diablo Canyon kjse/4328 Enclosures cc: Edgar Bailey, DHS Bruce S. Mallett David L. Prouix Diablo Distribution cc/enc: Girija S. Shukla A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

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PG&E Letter DCL-03-183 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Docket No. 50-275 In the Matter of ) Facility Operating License PACIFIC GAS AND ELECTRIC COMPANY) No. DPR-80

)

Diablo Canyon Power Plant ) Docket No. 50-323 Units 1 and 2 Facility Operating License No. DPR-82 I

AFFIDAVIT David H. Oatley, of lawful age, first being duly sworn upon oath says that he is Vice President and General Manager - Diablo Canyon of Pacific Gas and Electric Company; that he has executed license amendment request LAR 03-18 on behalf of said company with full power and authority to do so; that he is familiar with the content thereof; and that the facts stated therein are true and correct to the best of his knowledge, information, and belief.

David H. Oatley Vice President and General Manager - Diablo Canyon Subscribed and sworn to before me this 7th day of January 2004.

SANDRA EAMHERLY .I.

Notary Public Commission # 142556 Notory Public -California a County of San Luis Obispo San Luls Obtspo County State of California M Comm. ExpkresJgl 15. 20071

__~~~~~~~W111100 ,

Enclosure 1 PG&E Letter DCL-03-183 EVALUATION

1.0 DESCRIPTION

This letter is a request to amend Operating Licenses DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP), respectively.

The proposed changes would revise the DCPP Technical Specifications (TS) to allow application of a 4-volt alternate repair criteria (ARC) at intersections of steam generator (SG) tube hot-legs with the 4 lowest SG tube support plates (TSPs). The 4-volt ARC will only apply to Model 51 SG tubes experiencing outside diameter stress corrosion cracking (ODSCC) at the intersections of the tube hot-legs and the 4 lowest TSPs. The technical basis for the 4-volt ARC is described in Westinghouse Electric LLC WCAP-16170-NP, Revision 0, "Diablo Canyon SG Alternate Repair Criteria Based on Limited Tube Support Plate Displacement," dated November 2003 (nonproprietary), contained in Enclosure 4 of this letter and WCAP-16170-P, Revision 0, "Diablo Canyon SG Alternate Repair Criteria Based on Limited Tube Support Plate Displacement," dated November 2003 (proprietary), contained in Enclosure 6 of this letter.

In addition, the proposed change includes the application of leak-before-break (LBB) to the main steam line (MSL) piping inside containment in order to exclude the dynamic effects of a main steam line break (MSLB) in the short length of piping upstream of the MSL flow restrictor (large MSLB) from consideration for determining the loads on the SG TSPs following a MSLB. As required by General Design Criteria (GDC) 4 of Appendix A to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50), staff review and approval of the LBB analysis is required before a licensee can exclude the dynamic effects from the design basis. LBB will only be applied to limit the thermal hydraulic analysis to a small MSLB to determine the loads on the SG TSPs. The magnitude of the loads on the SG TSPs determines the number of SG tubes that need to be expanded into the TSPs to ensure limited TSP displacement following a postulated MSLB.

The 4-volt ARC is based on limited TSP displacement. The proposed LBB to the MSLs will only be applied until the DCPP SGs are replaced. The proposed change would not apply LBB to exclude other dynamic effects associated with postulated MSL pipe ruptures such as jet impingement, pipe whip, mass and energy release, etc. A summary report of the LBB analysis for the MSL is contained in Enclosure 7 to this letter.

2.0 PROPOSED CHANGE

The following changes are made to TS section 5.5.9, "Steam Generator (SG)

Tube Surveillance Program," to allow the 4-volt ARC:

Section 5.5.9.b.4 - new Section a) is added to the ODSCC inspection requirements to provide inspection requirements for bobbin voltages above the 1

Enclosure 1 PG&E Letter DCL-03-183 lower voltage repair criteria at TSP intersections other than at the 4 lowest hot-leg TSPs for SGs that are locked by expansion joints in accordance with WCAP-16170-P, Revision 0, methodology (locked SGs). New Section a) states; "For SGs that are locked by expansion joints in accordance with WCAP-1 6170-P, Revision 0, methodology (herein referred to as locked SGs), at tube support plate intersections other than at the four lowest hot-leg tube support plates, tubes with degradation attributed to axially oriented ODSCC within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (defined in 5.5.9.d.1.j) shall be inspected by Plus Point coil,"

Section 5.5.9.b.4 - new Section b) is added to the ODSCC inspection requirements to provide eddy current inspection requirements for TSP intersections at the 4 lowest hot-leg TSPs for locked SGs. New Section b) states; "For locked SGs at tube support plate intersections at the four lowest hot-leg tube support plates, tubes with degradation attributed to axially oriented ODSCC within the bounds of the tube support plate with a bobbin voltage greater than 4 volts shall be inspected by Plus Point coil. An additional 100 tube intersections at the four lowest hot-leg tube support plates with degradation attributed to axially oriented ODSCC within the bounds of the tube support plate with a bobbin voltage less than 4 volts (100 total of all SGs, not necessarily selected at random) shall be inspected by Plus Point coil,"

Section 5.5.9.b.4 - new Section c) is added to the ODSCC inspection requirements to provide the inspection requirements following the expansion process for expanded tubes. New Section c) states; "In each locked SG following the expansion process, 100% of the expanded tubes shall be Plus Point coil inspected at the top of tubesheet,"

Section 5.5.9.b.4 - new Section d) is added to the ODSCC inspection requirements to provide the inspection requirements for expanded tubes following two complete cycles of operation if the SGs are not replaced. New Section d) states; "In each locked SG, following two complete cycles of operation and if the SGs are not to be replaced at the outage, 20% of the expanded tubes shall be Plus Point coil inspected at the top of tubesheet and expanded tube support plate locations. The Plus Point coil inspection shall be expanded to 100% of the expanded tubes in any SG that a circumferential indication greater than 100 degrees is found at the top of the tubesheet or above or below the tube support plate edges at tube support plate locations,"

Section 5.5.9.d.1.j (iii) - A new sentence is added to the ODSCC TSP plugging limit for voltages between the lower and upper voltage repair limit to exclude TSP intersections at the 4 lowest hot-leg TSPs in locked SGs. The new sentence states; "This specification is not applicable to tube support plate intersections at the four lowest hot-leg tube support plates in locked SGs,"

2

Enclosure 1 PG&E Letter DCL-03-183 Section 5.5.9.d.1.j (v) - A new sentence is added to the ODSCC TSP plugging limit unscheduled mid-cycle inspection requirements to exclude TSP intersections at the 4 lowest hot-leg TSPs in locked SGs. The new sentence states; "This specification is not applicable to tube support plate intersections at the four lowest hot-leg tube support plates in locked SGs,"

Section 5.5.9.d.1.j (v) - A new sentence is added to the ODSCC TSP plugging limit lower voltage repair limit definition in Note 1 to provide the repair limit for TSP intersections at the 4 lowest hot-leg TSPs in locked SGs. The new sentence states; "For locked SGs, the lower voltage repair limit is 4.0 volts at tube support plate intersections at the four lowest hot-leg tube support plates and 2.0 volts at all other tube support plate intersections,"

Section 5.5.9.d.1 .1(v) - A new sentence is added to the ODSCC TSP plugging limit upper voltage repair limit definition in Note 2 to exclude the upper voltage repair limit from the 4 lowest hot-leg TSP intersections in locked SGs. The new sentence states; "For locked SGs, the upper voltage repair limit is not applicable at tube support plate intersections at the four lowest hot-leg tube support plates."

The following changes are made to TS Section 5.6.10, "Steam Generator (SG)

Tube Inspection Report":

Section 5.6.10.d.1 - A new sentence is added to the ODSCC ARC NRC notification requirements for leakage to provide leakage notification requirements for TSP intersections at the 4 lowest hot-leg TSPs in locked SGs. The new sentence states; "For locked SGs, the estimated leakage from tube support plate intersections at the four lowest hot-leg tube support plates shall be calculated using the methodology of WCAP-1 61 70-P, Revision 0, and combined with leakage from all other tube support plate intersections as calculated using the methodology of Generic Letter 95-05 modified to include a large leak rate assigned to indications predicted to burst in a Monte Carlo simulation as described in WCAP-16170-P, Revision 0."

Section 5.6.10.d.5 - A new sentence is added to the ODSCC ARC NRC notification requirements for burst probability to provide burst probability calculation requirements for TSP intersections at the 4 lowest hot-leg TSPs in locked SGs. The new sentence states; "For locked SGs, the burst probability is calculated using the methodology of WCAP-161 70-P, Revision 0, that is, the axial tensile tearing rupture probability at tube support plate intersections at the four lowest hot-leg tube support plates is combined with the conditional burst probability at all other tube support plate intersections,"

Section 5.6.10.d - A new Section 6 is added to the ODSCC ARC NRC notification requirements to provide notification requirements for support plate ligament cracking in expanded tubes. The new Section 6 states; "For locked SGs, if a 3

Enclosure 1 PG&E Letter DCL-03-183 significant increase in tube support plate ligament cracking is found in expanded tubes, an evaluation will be performed and the results reported to the NRC prior to restart,"

Section 5.6.10.d - A new Section 7 is added to the ODSCC ARC NRC notification requirements to provide notification requirements for circumferential indications in expanded tubes. The new Section 7 states; "For locked SGs, if circumferential indications are found in expanded tubes, an evaluation will be performed and the results reported to the NRC prior to restart."

In addition, LBB is applied to the MSL piping inside containment in order to exclude the dynamic effects of large MSLBs from consideration for determining the loads on the SG TSPs following a MSLB.

In summary, the proposed changes to TS Sections 5.5.9 and 5.6.10 support a new 4-volt axial ODSCC ARC at intersections of the tube hot-legs and the 4 lowest TSPs for SGs that are locked by expansion joints in accordance with WCAP-16170-P, Revision 0, methodology.

The DCPP TS Section 5.0, "Administrative Controls," does not have a bases section and therefore no changes to the TS Bases are required.

The proposed TS changes are noted on the markup TS page provided in Enclosure 2. The proposed retyped TS are provided in Enclosure 3.

3.0 BACKGROUND

3.1 SG Design The DCPP SGs are a Westinghouse Model 51 design. The SGs are vertical shell and U-tube evaporators with integral moisture separating equipment. The reactor coolant flows through inverted U-tubes, entering and leaving through the nozzles located in the hemispherical bottom head of the SG. Steam is generated on the shell side and flows upward through the moisture separators to the outlet nozzle at the top of the vessel. Each SG contains 3388 mill-annealed Alloy 600 U-tubes, 0.875 (7/8) inch outer diameter (OD), 0.050 inch wall thickness, which provide 51,500 ft2 of heat transfer area per SG. The SGs are discussed in detail in Chapter 5.5.2 of the DCPP Final Safety Analysis Report Update (FSARU).

The DCPP SGs utilize seven 0.75 inch thick carbon steel TSPs. The TSPs are numbered from 1 at the lowest elevation to 7 at the highest elevation and the lowest 4 TSPs are referred to as TSPs 1-4. The TSPs have drilled round tube holes and flow circulation holes set on a square pitch. The TSPs are supported vertically using a central tierod/spacer and 4 outer tierods/spacers located near the edge of the plate. The in-plane 4

Enclosure 1 PG&E Letter DCL-03-183 support for the TSPs is provided by six tapered wedges located around the periphery of each plate. The wedges are welded to both the wrapper and the TSP on both the top and bottom surfaces. As a result of the wedges being welded to both the TSP and wrapper, the wedges also provide restraint to vertical motions of the TSPs. There are also two bar supports welded between each TSP and wrapper, located 180 degrees apart.

The linear portions of the inverted SG U-tubes pass through TSPs at various levels which provide lateral support to the tubes. The TSPs contain circulation holes through which water/steam passes upward along the tube bundle. During normal operation, a slight pressure drop exists across each TSP. If the tube to TSP crevices are assumed to be open, this pressure drop would cause a small displacement of the TSPs relative to the tubes during normal operating conditions. However, the DCPP SGs have dented and packed crevices, and the TSPs would not be displaced relative to the cold shutdown or hot standby condition under normal operating conditions. At hot standby conditions, there is no secondary flow or pressure drop across the TSPs, and the hot standby positions are the same as the full power and cold shutdown TSP positions relative to the tube due to the crevice conditions. During postulated accident conditions, such as a MSLB, pressure differentials across individual TSPs are assumed to displace the TSPs. However, if sufficient SG tubes are expanded (locked) in the region of the TSPs, the displacement of the TSPs can be limited such that the uncovering of degraded tube spans, normally circumferentially constrained by their respective TSP, can be prevented.

Each MSL connected to the SG outlet nozzle contains a flow restrictor to limit the steam blowdown rate from the SGs in the event of a MSLB downstream of the flow restrictor. The MSL flow restrictors are described in Section 5.5.4 of the DCPP FSARU. In the event of a MSLB, steam flow rate is restricted in the throat of a venturi nozzle by a critical flow phenomenon whereby maximum flow corresponds to sonic velocity under the conditions in the throat. Each restrictor assembly includes a 16-inch Inconel throat venturi-nozzle section, and a carbon steel discharge cone welded to the throat. The complete restrictor assembly is fitted inside a length of main steam pipe and attached to the pipe by a circumferential weld at the discharge end. Materials, welding, and inspection requirements applied in fabrication of the restrictors conform to ANSI B31.1 and Section I of 1968 version of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requirements. The flow restrictor is located as close as feasible to the SG outlet nozzle to minimize the length of piping preceding the restrictor, thereby reducing the probability of a large MSLB. The design criterion for the MSL flow restrictors is to limit steam flow in the event of a MSLB 5

Enclosure 1 PG&E Letter DCL-03-183 downstream of the flow restrictor to 3,297 pounds mass per second at a steam pressure of 1,100 pounds per square inch to reduce the probability of fuel cladding damage.

3.2 SG Tube Integrity The SG tubes constitute more than half of the reactor coolant pressure boundary (RCPB). Design of the RCPB for structural and leakage integrity is a requirement under 10 CFR 50 Appendix A. Specific requirements governing the maintenance and inspection of SG tube integrity are in the DCPP TS, Section Xl of the ASME Boiler and Pressure Vessel Code, and Regulatory Guide (RG) 1.83. These include requirements for periodic inservice inspection of the tubing, flaw acceptance criteria (i.e., repair limits for plugging), and primary-to-secondary leakage limits. These requirements, coupled with the broad scope of plant operational and maintenance programs, have formed the basis for assuring adequate SG tube integrity.

SG tube plugging limits are specified in the DCPP TS. The current DCPP TS require that a tube with an imperfection be removed from service by plugging if the imperfection depth in the tube is greater than or equal to 40 percent through-wall, unless the degradation is subject to voltage-based ODSCC ARC, W* ARC, or primary water stress corrosion cracking (PWSCC) within dented TSP ARC.

TS 5.5.9 repair limits ensure that tubes accepted for continued service will retain adequate structural and leakage integrity during normal operating, transient, and postulated accident conditions, consistent with GDC 14, 15, 30, 31, and 32 of 10 CFR 50, Appendix A. Structural integrity refers to maintaining adequate margins against gross failure, rupture, and collapse of the SG tubing. Leakage integrity refers to limiting primary-to-secondary leakage to within acceptable limits.

Section 5.5.2.5 of the DCPP FSARU discusses the SG tube surveillance program including the inservice inspection requirements, primary-to-secondary leakage requirements, and SG ARC. Sections 15.3.1 and 15.4.1 of the DCPP FSARU discuss the small break loss-of-coolant accident (LOCA) and large break LOCA respectively. The small and large break LOCA analyses assumed 15 percent SG tube plugging in any SG, the maximum allowable SG tube plugging limit as indicated in DCPP FSARU Section 15.3.1.3.1 and FSARU Table 15.4-7A.

The consequences associated with a SG tube rupture (SGTR) event are discussed in DCPP FSARU Section 15.4.3. The limit for accident induced leakage through SG tubes is based on the FSARU Section 15.5.18.1 MSLB radiological consequences analysis. The current limit on postaccident induced leakage is 10.5 gallons per minute.

6

Enclosure I PG&E Letter DCL-03-183 3.3 ODSCC SG ARC The generic criteria for voltage-based limits for ODSCC are contained in GL 95-05. The generic criteria for voltage-based limits rely on empirically derived correlations between a nondestructive inspection parameter, the bobbin coil voltage, and tube burst pressure and leak rate. The GL guidance ensures structural and leakage integrity continue to be maintained at acceptable levels consistent with the requirements of 10 CFR 50 and the guideline values in 10 CFR 100 through augmented SG tube inspections and more restrictive operational leakage limits. The GL 95-05 repair limit is 1 volt for 0.75-inch OD tubing and 2 volts for 0.875-inch OD tubing.

GL 95-05 focuses on maintaining tube structural integrity during the full range of normal, transient, and postulated accident conditions with adequate allowance for eddy current test uncertainty and flaw growth projected to occur during the next operating cycle. In order to ensure the structural and leakage integrity of the tube until the next scheduled inspection, GL 95-05 specifies a methodology to determine the conditional burst probability and the total primary-to-secondary leak rate from an affected SG during a postulated MSLB event. The methodology in WCAP-14277, Revision 1, USLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections," dated December 1996, is used to implement the GL 95-05 structural integrity methodology.

The voltage-based ODSCC repair criteria are briefly described in FSARU section 5.5.2.5.4, "Voltage-Based Alternate Repair Criteria." The use of a voltage-based ARC for ODSCC indications at SG TSP intersections was approved by the NRC in Amendment Nos. 124 and 122 to Facility Operating License Nos. DPR-80 and DPR-82, respectively, for DCPP Units 1 and 2 in a letter to PG&E dated March 12,1998. PG&E requested the use of the voltage-based ARC for ODSCC at SG TSP intersections in PG&E Letter DCL-97-034, "License Amendment Request 97-03, Voltage-Based Alternate Steam Generator Tube Repair Limit for Outside Diameter Stress Corrosion Cracking at Tube Support Plate Intersections,"

dated February 26, 1997.

3.4 LBB GDC 4 of Appendix A to 10 CFR 50 allows the use of fracture mechanics methodology to exclude from structural design consideration the dynamic effects of pipe ruptures in nuclear power plants, provided it is demonstrated that the probability of pipe rupture is extremely low under conditions consistent with the design basis of the piping. The demonstration of low probability of pipe rupture utilizes a deterministic 7

Enclosure 1 PG&E Letter DCL-03-183 fracture mechanics analysis that evaluates the stability of postulated, small, through-wall flaws in piping and the ability to detect leakage through the flaws before the flaw could grow to unstable sizes and break the pipe.

This concept is referred to as LBB. The limitations and acceptance criteria for LBB are contained in Volume 3 of NUREG-1 061, "Report of the U.S.

Nuclear Regulatory Commission Piping Review Committee," dated April 1985.

In GDC 4, the NRC states, in part, that "dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis of the piping."

Therefore, staff review and approval of the LBB analysis is required before a licensee can exclude the dynamic effects from the design basis.

As discussed in NUREG-0800, Section 3.6.3, "Leak-Before-Break Evaluation Procedures," dated March 1987, the approval of LBB analyses by the staff permits the case-by-case removal of protective hardware such as pipe whip restraints and jet impingement barriers, the redesign of pipe connected components, their supports and their internals, and other related changes in operating plants. Expanding the SG tubes into the TSPs to achieve TSP 'locking" is considered to be a redesign of the pipe connected component internals. Thus application of LBB to exclude the dynamic effects of a large MSLB on the SG TSPs is appropriate.

3.5 Purpose for Proposed Amendments The DCPP SGs have experienced tube degradation related to axial ODSCC. SG tubes that experience excessive degradation reduce the integrity of the primary-to-secondary pressure boundary. Eddy current examination is used to measure the extent of SG tube degradation.

When the reduction in tube wall thickness reaches 40 percent of nominal SG tube wall thickness, the SG tube is considered defective and is removed from service by installing plugs in the SG tube at the inlet and outlet of the SG tube unless the tube can be left in service under an ARC. The DCPP TS contain a voltage-based axial ODSCC ARC which allows axial ODSCC degradation confined within the thickness of the SG TSP with a bobbin coil probe voltage less than or equal to 2 volts to remain in service. The 2-volt bobbin coil probe repair limit is based on the GL 95-05 allowable repair limit for 0.875-inch OD tubing.

The GL 95-05 2-volt repair limit was established to maintain the probability of tube burst less than 0.01 assuming design basis MSLB events may vertically displace a TSP. The displacement of the TSP will remove the 8

Enclosure 1 PG&E Letter DCL-03-183 constraint on the tube spans within the TSP and potentially allow degradation in the tube span to burst. However, for TSPs at hot-leg SG tube locations in which sufficient SG tubes have been expanded such that the TSP is locked in place, the deflections of the TSP following a limiting MSLB event are negligible. Therefore, the TSPs remain essentially stationary during all conditions, and the SG tube spans within the drilled TSP holes are constrained. For the intersections of the SG tube hot-legs and these TSPs, axial tube burst is eliminated as a credible event and a larger bobbin voltage can be allowed while still meeting the tube structural requirements of Regulatory Guide 1.121.

For TSPs which have been locked in place at hot-leg SG tube locations, the displacement of the TSPs following a design basis MSLB event is negligible and the tube burst probability is several orders of magnitude less than the 0.01 probability-of-burst criterion specified by GL 95-05.

Thus for axial ODSCC within locked TSPs, repair limits to preclude burst are not needed and tube repair limits may be based primarily on limiting leakage to acceptable levels during accident conditions.

Currently, axial ODSCC degradation confined within the thickness of the SG TSP with a bobbin voltage greater than 2 volts is removed from service by plugging the affected SG tube on both ends. However, the installation of plugs in a SG tube eliminates the heat transfer surface associated with the tube. In addition, plug installation leads to the reduction in the primary coolant flow available for core cooling. Therefore, the installation of plugs into SG tubes is currently limited to a maximum of 15 percent of the tubes in any SG.

At DCPP, a 4-volt axial ODSCC ARC will be supported by locking the 4 lowest TSPs in place. The use of a 4-volt axial ODSCC ARC in the SG tube hot-leg region of the 4 lowest TSPs will allow SG tubes with bobbin indications up to 4 volts within the 4 lowest TSPs to remain in service and thus minimize the reduction in SG tube heat transfer surface area and primary coolant flow due to tube plugging.

PG&E is currently planning on replacing the DCPP Unit 2 SGs in the fourteenth Unit 2 refueling outage, scheduled in 2008, and replacing the DCPP Unit 1 SGs in the fifteenth Unit 1 refueling outage, scheduled in 2009. The proposed 4-volt ARC and proposed application of LBB to the MSL piping to exclude large MSLBs will be used until the SGs are replaced.

PG&E requests approval of this LAR by October 1, 2004, to support the upcoming twelfth DCPP Unit 2 refueling outage (2R12) currently scheduled for October, 2004. Approval of this change will reduce the number of SG tubes which need to be removed from service by plugging 9

Enclosure 1 PG&E Letter DCL-03-183 during 2R12 and will also allow certain SG tubes which are currently plugged to be returned to service by removing the plugs.

4.0 TECHNICAL ANALYSIS

4.1 Application of LBB to Support Calculation of TSP Hydraulic Loads LBB is applied in order to exclude large MSLBs from consideration for determining the loads on the SG TSPs following a MSLB. Expanding the SG tubes into the TSPs is considered a redesign of the pipe connected component internals. As discussed in NUREG-0800, Section 3.6.3, "Leak-Before-Break Evaluation Procedures," dated March 1987, the NRC approval of LBB permits the application of such approval to the redesign of pipe connected component internals. LBB will only be applied to the main steam piping for determination of the loads on the SG TSPs to support the 4-volt ARC and is only needed until the DCPP SGs are replaced. LBB is not being used to permit removal of protective hardware such as pipe whip restraints and jet impingement barriers from the MSL, or to redesign the pipe connected component supports associated with the MSL. The details of the LBB analysis for the MSL are contained in Enclosure 7 to this letter.

The approach used to perform the LBB analysis for the MSLs is consistent with that used for LBB analyses approved by staff for the R.E. Ginna Nuclear Power Plant, the Millstone Nuclear Power Station Unit 2, and the Kewaunee Nuclear Power Plant: The LBB analyses for these plants were approved by staff in letters for Docket No. 50-244, "Staff Review of the Submittal by Rochester Gas & Electric to Apply Leak Before Break Status to portions of R. E. Ginna Nuclear Power Plant Residual Heat Removal System Piping (TAC No. MA0389)," dated February 25, 1999, Docket No. 50-336, "Staff Review of the Submittal by Northeast Nuclear Energy Company to Apply Leak Before Break Status to the Pressurizer Surge Line, Millstone Nuclear Power Station Unit 2 (TAC No. MA4146)," dated May 4, 1999, and Docket No. 50-305, "Kewaunee Nuclear Power Plant -

Review of the Leak Before Break Evaluation for the Residual Heat Removal, Accumulator Injection Line, and Safety Injection System (TAC No. MB1301)," dated September 5, 2002, respectively.

LBB Requirements GDC 4 of Appendix A to 10 CFR 50 allows the use of fracture mechanics methodology to exclude from structural design consideration the dynamic effects of pipe ruptures in nuclear power plants, provided it is demonstrated that the probability of pipe rupture is extremely low under conditions consistent with the design basis of the piping. The limitations and acceptance criteria for LBB are contained in NUREG-1061, Volume 3.

10

Enclosure 1 PG&E Letter DCL-03-183 LBB Analysis Summary The following summary of the LBB analysis is formatted according to the "Recommendations for Application of the LBB Approach" contained in the executive summary of NUREG-1061 Volume 3:

(a) The main steam piping systems are constructed of SA-516, Grade 70 carbon steel piping. At the operating temperature of 519 0 F, this material is very ductile and it is not susceptible to cleavage-type fracture. In addition, it has been shown that these systems are not susceptible to the effects of corrosion, high cycle fatigue or water hammer.

(b) Loadings have been determined from the original piping analysis, and are based upon pressure, dead weight, thermal expansion and safe shutdown earthquake (SSE) seismic motion and other dynamic loads such as main steam isolation valve (MSIV) rapid closure loads. Three seismic cases were considered in the DCPP design basis, the design earthquake, the double design earthquake, and the Hosgri earthquake. The highest of these, the Hosgri earthquake, was considered in the evaluation as the SSE case. The Hosgri case was enveloped with the MSIV loads (greater of the Hosgri or MSIV load was used). All stress locations in the main steam piping inside containment were considered.

(c) Plant-specific certified material test report (CMTR) data was used to establish conservative lower bound stress-strain properties to be used in the evaluations. For the fracture toughness properties, lower-bound generic industry material properties for the piping and welds were conservatively used in the evaluations.

(d) Crack growth analysis was conducted at the most critical locations on all the evaluated piping, considering the cyclic stresses predicted to occur over the life of the plant. For a hypothetical flaw with aspect ratio of 10:1 and an initial flaw depth of approximately 15 percent of pipe wall, the final flaw size after considering all plant transients for 40-year plant life is 41 percent flaw depth-to-thickness ratio, which is significantly less than ASME Code Section Xl allowable flaw size of 75 percent flaw depth-to-thickness ratio.

Hence, fatigue crack growth is not a concern for the main steam piping.

(e) Based on evaluation of all weld locations in the piping system in both DCPP Unit 1 and Unit 2, including axial welds and elbows, it was determined that the leakage from the crack at the limiting 11

Enclosure 1 PG&E Letter DCL-03-183 location was 2 gpm. NUREG-1061 Vol. 3 recommends that the leakage detection system be capable of measuring leakage of one tenth of this amount or 0.2 gpm.

(f) Each of the MSLs considered in the evaluation is approximately 90 feet in length and is not geometrically complex. All other dynamic loads that could occur in the systems, such as MSIV actuations, were considered with the SSE loads to determine the dynamic loads to be used in the evaluation.

(g) Crack growth of a leakage size crack in the length direction due to a dynamic event is no more than 1 percent of the leakage flaw size.

This is not significant compared to the margin between the leakage-size and the critical crack size.

(h) For all locations, the critical size circumferential crack was determined for the combination of normal plus SSE loads or MSIV loads, whichever is greater. The leakage size flaw was chosen such that its length was no greater than the critical crack size reduced by a factor of two. Axial cracks were considered and were shown to exhibit much higher leakage and more margin than circumferentially oriented cracks.

(i) For all locations, the critical crack size was determined for the combination of the square root of two times the normal plus dynamic loads. The leakage size crack was selected to be no greater than this critical crack size. (The minimum of the crack sizes determined by this criterion, and that of the criterion of (h) above, was chosen for calculation of the leakage rate for each location.)

(O-n) No special testing (other than information in the CMTRs) was conducted to determine material properties for fracture mechanics evaluation. Instead, information from the piping CMTRs was used to derive lower bound material toughness and tensile properties in the evaluations. The material properties so determined have been shown to be applicable near the upper range of normal plant operating conditions and exhibit ductile behavior at these temperatures.

(o) Limit load analysis was not utilized in the LBB evaluation since the main steam piping material is carbon steel. The elastic-plastic fracture mechanics J-integrallTearing Modulus analysis approach was used to determine the critical flaw sizes.

12

Enclosure 1 PG&E Letter DCL-03-183 Conservatisms in LBB analysis The LBB analysis for the DCPP MSL piping included the following conservative assumptions which are in addition to the conservative assumptions required by NUREG-1061, Vol. 3, and NUREG-0800, Section 3.6.3:

1) The SSE loads are based on a conservative envelope of building response spectra from multiple elevations. The highest seismic loads are calculated to occur at the elevation with the lowest acceleration input. Use of the multiple level spectral method would likely result in reduced seismic loads at the limiting location. Lower seismic loads would result in larger critical flaw sizes and therefore, larger leak rates.

2) The MSIV loads used in the evaluation were conservatively calculated.

The loads were developed nearly twenty-five years ago using limited computation capabilities. Use of currently available software and methodologies, along with more realistic assumptions, would likely reduce these loads. Reduced loads will result in larger critical flaw sizes and therefore, higher leakage rates.

3) Lower bound ASME Code minimum material properties were used for the stress-strain parameters in the fracture mechanics evaluation.

4) Lower bound fracture toughness properties, which formed the basis for the flaw acceptance in ASME Code Section Xl, was used to determine the critical flaw sizes.

The requirements of NUREG-1061, Volume 3, and NUREG-0800, Section 3.6.3 provide significant conservatism to the LBB evaluation.

There is a requirement of a safety factor of two between the critical flaw size and the leakage flaw size. If this safety factor is reduced to 1.5, the calculated leakage will increase considerably. Reduction of this safety factor will still allow a significant amount of time to detect leakage before the critical flaw size is reached, since crack growth is minimal.

It is concluded, using the requirements of NUREG-1061, Volume 3, that the MSL piping at DCPP qualifies for the application of LBB to exclude the dynamic effects of a MSLB upstream of the MSL flow restrictor from consideration for determining the loads on the SG TSPs following a MSLB. The limiting leakage was determined to be 2 gpm. The inspections to support LBB for the MSLs and the capability of the leakage detection system at DCPP to detect a 0.2 gpm steam leak from a MSL are addressed below.

13

Enclosure 1 PG&E Letter DCL-03-183 Inspections to Support LBB for MSLs Currently, per the Risk Informed Inservice Inspection Program, all MSL welds are low risk and hence are not selected for inspection. However, to support LBB for the MSLs, the limiting welds in terms of stress and leak detection of the main steam line inside containment will be inspected. The limiting weld in terms of stress and leak detection for the MSLs is the pipe-to-flued head weld at the containment penetration. While LBB is credited for the MSLs, the limiting welds for the MSLs will be inspected at a frequency described by ASME Section Xl Inservice inspection requirements, equally divided over three inservice inspection periods. A baseline inservice inspection will be conducted on these welds prior to crediting LBB for the MSLs.

Leak Detection System Capability to Support LBB for MSLs Detailed information on the reactor coolant system (RCS) leakage detection system is provided in Section 5.2.7 and Table 5.2-16 of the DCPP FSARU. The nonradioactive leak detection capabilities are discussed below. The containment condensation measuring system provides a measure of the amount of leakage vaporized. This system collects and measures moisture condensed from the containment atmosphere by the cooling coils of the fan cooler air circulation units. This system dependably and accurately measures total vaporized leakage.

Additional or abnormal leaks will result in containment humidity and condensation runoff rate increases, and the additional leakage can then be measured. Section 5.2.7.4 of the DCPP FSARU states; "A preliminary estimate of the evaporated leakage can be obtained from the condensate flow increase rate during the transient; a better estimate can be determined from the steady state condensate flow when equilibrium has been reached. After equilibrium is attained, condensate flow from approximately 0.1 to 30 gpm per detector can be measured by this system." Various air temperature and pressure sensors are used to supplement indications of leakage into containment. A rise above the normally indicated range of values may indicate leakage into the containment atmosphere. Leakage from the primary system would result in reactor coolant flowing into one of the containment sumps. Sump level and sump pump integrated flow is monitored to provide a measure of the overall leakage that remains in liquid state.

The LBB analysis for the MSLs determined that the leakage at the limiting location was 2 gpm. NUREG-1061 Vol. 3 recommends that the leakage detection system be capable of measuring leakage of one tenth of this amount or 0.2 gpm. Detection of a 0.2 gpm steam leak from the MSL will require a more efficient monitoring method than currently credited in the FSARU. As such, data was gathered in November 2003 to determine the 14

Enclosure 1 PG&E Letter DCL-03-183 sensitivity of the level indicators for the two containment structure sumps.

These level indicators are used to determine total leakage into containment. A 0.3 inch change (0.005 volt) in sump level is discernible with the level indicators Ll-60 and LI-61, for the DCPP Unit I Structure Sump 1-1 (2-1 for DCPP Unit 2) and Structure Sump 1-2 (2-2 for DCPP Unit 2), respectively. The readability of this signal, in conjunction with the fact that the containment is effectively a closed control volume will allow the sumps to be used to detect a steam leak rate that would result in the collection of approximately five gallons in either sump. Since operation with various combinations of fan coolers in service may occur, it is assumed that half of the total leakage collects in each sump. For example, for a one hour observation period, it would be reasonable to expect a leak rate of 0.2 gpm (approximately ten gallons per hour) to be indicated by a five gallon change in each sump. Nonuniform fan cooler alignment to the sumps will result in a more rapid accumulation and detection (less than one hour) in one sump. A design change will be completed as part of the implementation to use the 4-volt ARC to allow the plant process computer (PPC) to monitor the structure sump levels continuously and to provide appropriate indication to the operators in the event of a 0.2 gpm MSL leak.

The NRC has previously approved application of LBB based on a 0.25 gpm leak rate detection capability for RCS leakage for Nuclear Management Company LLC's Kewaunee Nuclear Power Plant in the letter "Kewaunee Nuclear Power Plant - Review of Leak-Before-Break Evaluation for the Residual Heat Removal, Accumulator Injection Line, and Safety Injection System (TAC No. MB1301)," dated September 5, 2002, and for Rochester Gas and Electric Company's R. E. Ginna Nuclear Power Plant in the letter "Staff Review of the Submittal by Rochester Gas and Electric Company to Apply Leak-Before-Break Status to Portions of the R.E. Ginna Nuclear Power Plant Residual Heat Removal System Piping (TAC No. MA0389)," dated February 25, 1999.

4.2 Thermal Hydraulic Evaluation of Loads on the TSPs Thermal-hydraulic loads on the TSPs for application to the TSP displacement analyses were obtained using the RELAP5 code. The RELAP5 code has previously been used to calculate thermal hydraulic loads on TSPs of SGs during MSLB transients for DCPP as described in Westinghouse WCAP-14707, "Model 51 Steam Generator Limited Tube Support Plate Displacement Analysis for Dented or Packed Tube-to-Tube Support Plate Crevices," dated August 1996, which was submitted to the NRC in PG&E Letter DCL-96-206 dated October 4, 1996. The RELAP5 code has also previously been used to calculate thermal hydraulic loads on TSPs of SGs during MSLB transients and feedline break transients for the Braidwood Nuclear Power Station Unit 1 (Braidwood-1) as 15

Enclosure 1 PG&E Letter DCL-03-183 documented in Westinghouse WCAP-14273, "Technical Support for Alternate Plugging Criteria with Tube Expansion at Tube Support Plate Intersections for Braidwood-1 and Byron-1 Model D4 Steam Generators,"

dated February 1995, and for the South Texas Project Unit 2 as documented in Westinghouse WCAP-15163, Revision 1, "South Texas Project Unit 2; 3V Alternate Repair Criteria Application of Bounding Analysis and Tube Expansions," dated January 2001.

The RELAP5 modeling and analysis of the SGs are discussed in Section 4.0 of WCAP-16170-P, Revision 0. The RELAP5 code TSP hydraulic loads bound those obtained using the TRAC-M code for a Westinghouse Model 51 SG based on a comparison to TRAC-M hydraulic loads reported in the draft NUREG report "Sensitivity Studies of Failure of Steam Generator Tubes during Main Steam Line Break and Other Secondary Side Depressurization Events," dated December 2002. The analyses show that the TSP loads are higher for a MSLB event at hot standby operating conditions than for full power conditions. The hot standby loads were used for the tube expansion design even though only a small fraction of the operating cycle is spent at hot standby conditions.

Sensitivity analyses were performed to assess the dependence of the TSP loads on modeling and analysis uncertainties. It was determined that increasing the reference or expected case hot standby loads by a factor of 1.5 represents a conservative factor that combines the potential modeling uncertainties that could increase the TSP hydraulic loads over the reference case hot standby loads. This factor of 1.5 increase on the reference hot standby loads was applied to obtain the design basis TSP loads. The design bases TSP loads are needed to determine the number of tube expansions required to ensure limited TSP displacement. Thus, the hydraulic loads used are conservative.

4.3 Pressure Differential Across TSPs During SG Blowdown The effects of a MSLB will diminish with time as the SG depressurizes and the flow out the break decreases. As long as the pressure in the SG is high enough and the break large enough to result in choked flow at the flow restrictor, pressure fluctuations in the MSL downstream of the flow restrictor will not be able to propagate into the SG. If the area of the break is small enough (less than about 0.45 square feet or about one third of the area of the flow restrictor), the break flow will be less than that normally experienced during operation. The internals of the SG should not be significantly affected since there is considerable operating experience at this level of flow. Nevertheless, it may be possible that for a medium sized break, for which the break area is smaller than the nozzle area, the break flow could exceed the full power operating flow and the flow restrictor may not be choked. Under these conditions, pressure fluctuations in the MSL 16

Enclosure 1 PG&E Letter DCL-03-183 could possibly propagate into the SG and affect the internals.

Nevertheless, the significant change in area and the presence of the compressible steam in the large volume at the top of the SG combine to act as an accumulator and will help to isolate the lower internals from the effect of sudden pressure changes in the MSL.

Additional isolation for the tube bundle region is provided by the significant resistance that exists across the two levels of steam separators and the presence of large amounts of saturated liquid that can flash to maintain the pressure near saturation pressure. As a result, any sudden depressurization in the MSL leads to a much slower depressurization of the SG as a whole and relatively small pressure gradients would be expected inside the tube bundle. The pressure gradients that are established are primarily a result of "steady flow" rather than dynamic imbalance due to flow acceleration. The dominant loads on the TSPs in the tube bundle result from the swell of the fluid trapped by the TSPs as the SG begins to depressurize rather than from the propagation of sonic waves from the main steam nozzle.

To estimate the extent to which pressure fluctuations in the MSL could propagate into the tube bundle of the SG, a two-phase thermal-hydraulic analysis was conducted for which a sinusoidal pressure oscillation was imposed at the MSL boundary. The SG was assumed to be at hot standby conditions. The pressure response in the tube bundle region was determined as a function of the applied oscillatory pressure in the MSL.

The analyses were run until steady state oscillating conditions were achieved. Several such analyses were conducted using several different frequencies for the pressure oscillations to determine the frequency transform for the pressure oscillations between the MSL and the tube bundle region.

Results were obtained from 5 separate cases with pressure oscillation frequencies between 10 and 50 Hertz originated in the MSL of a Model E2 SG, which is similar to the Model 51 SG. These results provide the relative amplitude of the calculated response of pressure at the inside of the steam nozzle, at the top of the tube bundle, and at the region just above the tubesheet as compared to the amplitude of the pressure oscillations imposed at the MSL boundary. At low frequency, the calculated amplitude of the pressure oscillations at the tubesheet is about seven percent of the amplitude of the applied pressure oscillations in the MSL whereas the amplitude of the pressure oscillations at the U-bends is about two percent of the applied amplitude. There appears to be some frequency dependence for the response at low frequencies, particularly near the steam nozzle. This may indicate an acoustic resonance effect at the top of the SG since the response is about 90 degrees out of phase with the applied pressure. However, the response in the tube bundle 17

Enclosure 1 PG&E Letter DCL-03-183 remains low for all the analyzed frequencies. For frequencies above 30 Hertz, the calculated response in the tube bundle is negligible.

SG TSP Load Analysis Conclusions Hydraulic loads on the TSPs during the MSLB and feed line break have been considered. The hydraulic loads vary with different initial conditions.

Generally, SGs are in full power operation; however, they can be in the hot standby state without power generation. Therefore, both hot standby and full power were considered in the load analyses. Since the hydraulic loads can vary with various input parameters and with the uncertainty of those parameters, a sensitivity study was conducted to derive an uncertainty factor for providing conservative loads for input to the structural evaluation. Based upon the evaluation, the following conclusions were drawn:

1. Hydraulic loads on the TSPs during the hot standby initiated MSLB with both a small and large break calculated by the RELAP5 code, bound those obtained with the TRAC-M code.

2. Hydraulic loads on the TSPs during the hot standby initiated small MSLB bound those during the hot standby or full power initiated feed line break (FLB) for TSPs with loads large enough to be of concern for TSP displacement analyses (i.e., excluding the two lowest TSPs).

3. Hydraulic loads on the TSPs during the hot standby initiated MSLB bound those during full power initiated MSLB.

4. Hydraulic loads on the TSPs during the hot standby initiated MSLB with large or small break bound those due to hot standby or full power initiated FLB.

5. A 1.5 uncertainty factor can be used to multiply the best estimate hydraulic loads to conservatively bound the hydraulic loads calculated with worst conditions of the various input parameters.

For the 4 lowest TSPs at which the limited TSP displacement ARC is applied, the RELAP5 code small break loads are greater than a factor of 1.5 larger than the loads obtained with the TRAC-M Code. The factor of 1.5 applied to envelop analysis uncertainties provides additional conservatism for the hydraulic loads. The small break loads were calculated for a MSLB downstream of the MSL flow restrictor based on application of LBB to the MSL piping, which excludes large MSLBs.

4.4 SG TSP Displacement Analysis To obtain a negligible probability of burst (POB), an acceptable TSP displacement of less than or equal to 0.362 inch was very conservatively 18

Enclosure I PG&E Letter DCL-03-183 developed by assuming that the lowest 4 TSPs are uniformly displaced and by making the bounding assumption that these displacements exposed through-wall cracks at all hot-leg intersections at the 4 lowest TSPs (13,552 through-wall indications). At a uniform TSP displacement of 0.362 inch, the total POB for a SG at a primary-to-secondary pressure difference of 2405 pounds per square inch (psi) is less than 10-5, which is negligibly small compared to the conservative limit on POB of 10-2 given in NRC GL 95-05 for ODSCC at the TSPs. Assuming a factor of 1.5 on the transient loads calculated using RELAP5 to envelop modeling and calculation uncertainties, expansions in at least 30 tubes will result in maximum TSP displacements less than 0.15 inch for all tube locations at the hot-leg of the 4 lowest TSPs for a MSLB event initiated from hot standby conditions. At the goal TSP displacement of 0.15 inch, the POB is significantly less than 10 ° for a single tube. Section 8.0 of WCAP-1 6170-P, Revision 0, provides the basis for the POB evaluations.

High temperature leak rate tests were performed to determine the leak rates for indications restricted from burst (IRB). Section 7.0 of WCAP-16170-P, Revision 0, summarizes these tests. The bounding leak rate for the DCPP MSLB primary-to-secondary pressure differential of 2405 psi, based on the power operated relief valves (PORVs) for pressure relief, is 5.0 gpm. The bounding leak rate is based on tests with offsets (TSP displacements) up to 0.21 inch. The goal TSP displacement for the DCPP ARC was set at 0.15 inch to provide additional conservatism for the acceptable displacement of 0.21 inch for utilization of the IRB leak rate.

Due to the flexibility of the TSPs for out-of-plane loads, with no expanded tubes, the displacements of the plates may exceed 0.15 inch following a MSLB event. In order to limit the plate displacements, tubes will be expanded as described in WCAP-1 6170-P, Revision 0, such that the expansion length extends both above and below the TSPs. In this way the tubes will act as additional supports, retarding the vertical displacement of the plates. The number and location of the tube expansions varies from plate to plate depending on the magnitude of the applied pressure loading.

The TSP displacement analysis involved the preparation of a 180 degree tube bundle model that included both the hot-leg and cold-leg sides of the tube bundle, consisting of seven TSPs, tierods, spacers, channel head, lower shell, wrapper, and tubesheet. Elements were also included to represent the expanded tubes and the expansion zone interfaces between the tubes and support plates. The WECAN computer code, a general purpose finite element code, was used to develop the model. Model calculations were performed to define applicable dynamic degrees of freedom (DOF). Once the DOF were defined, a global substructure was generated for the overall tube bundle. The dynamic response of the 19

Enclosure 1 PG&E Letter DCL-03-183 plates was then calculated using the special purpose computer program, pltdym. The Staff previously reviewed and accepted the use of the computer code pltdym for predicting plate displacements under MSLB loads in the letter "NRC Staff Review of WCAP-14707/14708, 'Model 51 Steam Generator Limited Tube Support Plate Displacement Analysis for Dented or Packed Tube-to-Tube Support Plate Crevices' - Diablo Canyon Power Plant, Units 1 and 2 (TAC Nos. M99011 and M99012)," dated January 10, 2000.

The dynamic analysis was performed assuming elastic response of the plate support structures. Thus, in order to support the analysis TSP displacement results, calculations were performed to demonstrate the applicability of the elastic analysis approach in determining the resulting TSP displacements. These calculations demonstrated that the tierods/spacers remain elastic throughout the transient, that the expanded tubes remain elastic, that yielding (displacement greater than 0.15 inch) of the TSPs does not occur, and that the welds joining the wedges and support bars to the wrapper remain intact.

4.5 4-Volt ARC 4-Volt ARC TS Changes The proposed change increases the bobbin coil probe voltage limit to 4 volts in the TS 5.5.9 SG tube axial ODSCC ARC for degradation within the hot-leg region of the 4 lowest TSPs for a locked SG. New SG tube inspection requirements and acceptance criteria are added to TS 5.5.9.

Reporting requirements are added to TS 5.6.10 to support the 4-volt ARC.

The technical basis for the required TS changes for the 4-volt ARC is documented in Section 10.0 of WCAP-16170-P, Revision 0, contained in Enclosure 6. Changes are also made to the TS 5.5.9 and TS 5.6.10 existing 2-volt ARC wording to clarify the SG TSPs to which the 4-volt ARC will be applied. The technical bases for the TS changes are summarized below.

Section 5.5.9.b.4 A new section a) provides Plus Point inspection requirements, for bobbin voltages above the lower voltage repair criteria, at TSP intersections other than at the 4 lowest hot-leg TSPs for SGs that are locked by expansion joints in accordance with WCAP-16170-P, Revision 0, methodology (referred to as locked SGs). Required Plus Point coil inspection of bobbin indications greater than 2 volts provides additional inspection data for these indications.

Section 5.5.9.b.4 A new Section b) provides Plus Point coil inspection requirements for bobbin voltages above 4 volts and for 100 additional intersections below 4 volts at the TSP intersections at the 4 lowest hot-leg 20

Enclosure 1 PG&E Letter DCL-03-183 TSPs for locked SGs. Required Plus Point coil inspection of these bobbin indications provides additional inspection data for these indications.

Section 5.5.9.b.4 A new Section c) provides Plus Point coil inspection requirements following the expansion process for all the expanded tubes.

This ensures that circumferential cracks in the tubesheet region which could be created during the expansion process would be identified.

Section 5.5.9.b.4 A new Section d) provides the Plus Point coil inspection requirements for expanded tubes following two complete cycles of operation if the SGs are not replaced. Twenty percent of the expanded tubes shall be Plus Point inspected at the top of the tubesheet and expanded TSP locations to identify any cracking which could impact the ability of the expanded tubes to provide structural support to the TSPs. A 20 percent sample is considered an adequate sample size to identify any cracks which would exist in the expanded tubes and is consistent with the 20 percent sample size used to determine the lowest cold-leg TSP intersection having ODSCC indications in current TS 5.5.9.b.4. The inspections of the expanded tubes are not required prior to two complete cycles of operation based on the low likelihood of circumferential cracking at the expanded joints, that a severed parent tube within the TSP would have no significant effects on displacements, and that axial cracks at the expansions would not significantly influence the expanded joint stiffness.

The new TS 5.5.9.b.4 Section d) requires that the Plus Point coil inspection shall be expanded to 100 percent of the expanded tubes in any SG that a circumferential indication greater than 100 degrees is found at the top of the tubesheet or above or below the tube support plate edges at tube support plate locations. The 100 degree crack size is less than the predicted near through-wall length for undetected circumferential indications left in the expanded tubes after two cycles of operation assuming growth rates for active tubes. If circumferential indications are found in the expanded tubes, the potential need for corrective actions will be evaluated and the results of this evaluation reported to the NRC prior to restart. This evaluation and NRC reporting requirement is implemented through a new reporting requirement added to TS Section 5.6.10.d addressed below.

Section 5.5.9.d.lj (iii) A new sentence is added that prevents a voltage repair limit greater than 4 volts from being used for TSP intersections at the 4 lowest hot-leg TSPs. Four volts is the lower voltage repair limit for TSP intersections at the 4 lowest hot-leg TSPs irrespective of the rotating pancake coil confirmation. This change is conservative since a repair limit significantly greater than 4 volts could be justified to provide tube structural integrity for TSP intersections at the 4 lowest hot-leg TSPs.

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Enclosure 1 PG&E Letter DCL-03-183 Section 5.5.9.d.l.j (v) A new sentence is added to prevent the upper voltage repair limit from being applied to the TSP intersections at the 4 lowest hot-leg TSPs during unscheduled mid-cycle inspections and maintains the repair limit at 4 volts for these indications. This change is conservative since a repair limit greater than 4 volts could be justified for TSP intersections at the 4 lowest hot-leg TSPs.

Section 5.5.9.d.1.j (v) A new sentence is added to Note 1 which provides the 4-volt ARC for the TSP intersections at the 4 lowest hot-leg TSPs. For the TSP intersections at the 4 lowest hot-leg TSPs, a 4-volt repair limit is acceptable to meet tube burst criteria. The 4-volt repair limit is conservative since axial tensile rupture data support a much higher repair limit.

Section 5.5.9.d.1 .j (v) A new sentence is added to Note 2 to prevent the upper voltage repair limit from being applied to the TSP intersections at the 4 lowest hot-leg TSPs and maintains the repair limit at 4 volts for these indications. This change is conservative since a repair limit greater than 4 volts could be justified for TSP intersections at the 4 lowest hot-leg TSPs.

Section 5.6.10.d.1 A new sentence is added to include the revised leak rate requirements of WCAP-1 6170-P, Revision 0, which are a change to the requirements of GL 95-05, in the TS. The revised leak rate requirements will ensure that all potential leakage sources are considered for determination of the total leak rate.

Section 5.6.10.d.5 A new sentence is added to include the revised burst probability calculation requirements of WCAP-1 6170-P, Revision 0, which are a change to the requirements of GL 95-05. The revised burst probability requirements ensure that all burst mechanisms are considered at the TSP intersections.

Section 5.6.10.d A new Section 6 provides notification requirements for a significant increase in support plate ligament cracking in expanded tubes.

Consistent with the current TSP inspections performed to meet the requirements of TS 5.5.9.b.2.d and TS 5.5.9.b.4, the minimum eddy current inspection of TSP ligaments at each scheduled outage includes 100 percent bobbin coil with Plus Point confirmation for new bobbin calls and a 20 percent Plus Point inspection of TSP ligament indication previously confirmed by Plus Point inspections. The inspection of active tubes adjacent to expanded tubes shall include a bobbin coil examination for cracked TSP ligaments at expanded tube TSP intersections. If a cracked ligament indication is found and confirmed by Plus Point inspection as a cracked or missing ligament section, the potential implications on TSP displacements shall be assessed and documented in 22

Enclosure 1 PG&E Letter DCL-03-183 the ODSCC 90-day report. However, a significant increase in ligament cracking at the expanded TSP intersections, such as more than two indications at the same TSP elevation, shall be evaluated for implications on TSP displacement and this evaluation reported to the NRC prior to restart. The notification requirements for a significant increase in support plate ligament cracking in expanded tubes is incorporated into the new Section 6 to TS Section 5.6.10.d. This will ensure the NRC is notified, prior to restart, of the results of the evaluation of the effect of a significant increase in TSP ligament cracking on the SG TSP displacement analysis supporting the 4-volt ARC requirement.

Section 5.6.10.d A new section 7 provides notification requirements if circumferential indications are found in expanded tubes. This will ensure the NRC is notified, prior to restart, of the results of the evaluation of the impact of the circumferential indications on the 4-volt ARC.

SG Tube Deformation In addressing the potential combined LOCA and earthquake effects on SG components as required by GDC 2 of Appendix A to 10 CFR 50, analysis has shown that SG tube deformation may occur in certain regions of the SG. This deformation is caused by TSP plastic deformation in the region of the TSP wedge supports that can occur when TSPs experience large lateral loads concentrated at wedge support points on the periphery of a TSP due to the combined loading effects of a LOCA rarefaction wave and an earthquake. Deformation impinges on TSP apertures through which tubes pass, deflecting tube walls inward. The resulting pressure differential across deformed tube walls may cause some tubes to collapse.

SG tube collapse reduces RCS flow and could cause partial through-wall tube cracks to become full through-wall tube-cracks during tube deformation or collapse resulting in potential secondary-to-primary in-leakage to the RCS. Tubes for which deformation may occur are excluded from application of the voltage-based ARC per current TS 5.5.9.d.1 .j (iv). This exclusion will continue to apply for the 4-volt ARC and the exclusion is not adversely affected by the 4-volt ARC. Therefore, tubes for which deformation may occur will not be left in service under the 4-volt ARC.

Prior SG Internal Inspections SG internal inspections have been performed in preheater SGs at three plant sites to support limited TSP displacement ARCs. The inspections included extensive visual inspection of the top TSP in one SG at each site and no degradation was seen during the inspections. Visual inspection was performed on eight of the ten tierod nuts in one of the four SGs at two sites. The nuts were found to be tight against the top support plate with 23

Enclosure 1 PG&E Letter DCL-03-183 no degradation seen. Eddy current inspection was performed on all four SGs, at each site, in the area of the anti-rotational devices at the top TSP for identification of cracked TSP ligaments. In addition, eddy current inspection of the patch plate area in all four SGs, at each site, was performed and no degradation was noted. Approximately five wedges and eighty-nine support bars were visually inspected at each site and no degradation was noted. Also, no shift in the wrapper location was noted in any SG at the sites.

PG&E has performed secondary side inspections of DCPP Unit 1 and Unit 2 SGs during past refueling outages. The results of the inspections are summarized in Tables 9-1 and 9-2 of WCAP-16170-P, Revision 0.

No load path degradation was observed during any of these inspections although minor TSP cracking and loss of ligaments has been detected by eddy current. These inspections included routine refueling outage eddy current inspection of the carbon steel TSPs for ligament cracking for both Units, and visual inspection of the Unit 1 plates, wedges, tierods/spacers, support bars and wrappers during Unit 1 refueling outage 8 (1 R8).

The DCPP eddy current inspections have identified several hundred TSP ligament indications, mostly in Unit 1. The majority of the indications are traceable to preservice inspection, indicating they are not service-induced degradation. Some of the indications are associated with patch plate seam welds and others are associated with missing ligaments as verified by visual inspections conducted in 1R8. Westinghouse has concluded that the missing DCPP TSP ligaments observed during the 1R8 visual inspections are related to suspected TSP drilled hole manufacturing anomalies. The TSP manufacturing practices employed at the time that the DCPP SGs were produced used a stacked drilling procedure. Several TSPs were clamped together and drilled simultaneously. A review of the ligament indication locations indicated distinct location patterns, indicative of manufacturing anomalies of the automatic drilling equipment. The 1R8 visual examination of some of the locations also confirmed that they did not appear to be service-induced degradation. None of the ligament gap sizes have approached the 146 degree acceptance limit for tube plugging.

The largest gap to date is about 120 degrees based on Plus Point coil inspection. No tube wear at intersections with TSP ligament indications has been detected.

In summary, based on inspections performed to date in the domestic industry and at DCPP, no load path component degradation with the exception of TSP ligament cracking has been observed in Westinghouse fabricated SGs with denting the same or less than that found at DCPP.

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Enclosure 1 PG&E Letter DCL-03-183 SG Internal Inspection The 4-volt ARC for axial ODSCC is based on the indications being constrained within the TSP intersection and that TSP displacement is limited during a MSLB event. The thermal hydraulic analyses to determine the loads on the TSPs during normal operating and accident conditions assumes that certain SG internal components are present that provide a load path which is credited to limit TSP displacement. To ensure the load path components are not degraded to the level that they can no longer perform their intended function, a SG internals inspection program will be implemented. The program contains specific inspection logic, scope, inspection method, and acceptance criteria. This inspection plan is described in Section 9.6 of WCAP-16170-P, Revision 0.

To further assure the integrity of the load path components in the DCPP SGs, additional inspections through available access openings will be performed in upcoming refueling outages. The purpose of the inspections is to provide confidence that the structural load path required to support a limited TSP displacement ARC for the SGs is not degraded. The extent of these inspections will be based on existing accessibility, risk of "sticking" visual probes, radiation exposure, and outage schedule.

Continued eddy current inspections of TSPs will be performed in a manner consistent with prior inspections. The plan is to perform a 20 percent Plus Point inspection of existing baseline indications and 100 percent Plus Point inspection of all new bobbin suspect ligament crack indications. The term existing baseline indications includes all prior Plus Point confirmed TSP ligament crack indications and ligament gap indications. Marginal calls not confirmed in later Plus Point inspections may be excluded from the baseline indications. If an active degradation is detected by the 20 percent sample inspection, Plus Point inspection shall be expanded to a 100 percent baseline inspection. Active degradation is defined as service-induced TSP ligament erosion-corrosion and/or cracking.

Sample visual inspections will be performed in 1R12 in one SG as described in Table 9-5 of WCAP-1 6170-P, Revision 0. The visual inspections in the SG will include inspection at three TSPs of samples of the tierods/spacers, portions of TSPs, wedges, and support bars and welds. The visual inspections will also include a sample of tie rod nuts on top of the highest TSP (TSP 7).

The inspection plan also includes inspections to be performed during the outage in which TSP locking is implemented in both DCPP Unit 1 and Unit 2 and in each outage until SG replacement in all SGs as described in Table 9-4 of WCAP-16170-P, Revision 0. The inspection in all SGs will 25

Enclosure 1 PG&E Letter DCL-03-183 consist of a visual inspection of the tierods/spacers at the top of the tubesheet, a TSP eddy current inspection of 20 percent baseline indications and new indications with Plus Point coil and 100 percent bobbin coil, and a visual wrapper inspection in the lower shell and wrapper.

In summary, the inspection plan includes inspections for upcoming refueling outages, inspections to be performed during 1R12 in one SG, inspections to be performed in all SGs during the outage in which TSP locking is implemented, and inspections in each outage until SG replacement. Based on the results of previous SG internals inspections at DCPP, PG&E believes this inspection plan is sufficient to ensure long term integrity of the SG internal components and TSPs and to support a 4-volt ARC until SG replacement.

Tube Expansion Installation Procedures The tube expansions at the TSPs are performed by a hydraulic expansion process that expands the parent tube and the sleeve stabilizer at the same time. Expansions are performed below and above each TSP intersection that requires expansion. The expansion process will ensure the minimum expanded tube stiffness and minimum expanded tube diameter are met. Section 6.0 of WCAP-1 6170-P, Revision 0, describes the development tests performed to ensure the process meets the stiffness and diameter requirements. The sleeve stabilizer expanded with the parent tube increases the expansion stiffness at a given diametral expansion and prevents lateral motion or adjacent tube damage for a postulated severed expansion. The design target range for the expansion process leads to a low likelihood of circumferential cracks at the bulges.

Field application will be performed using a Framatome based sleeving system, which includes the Search and Locate End Effector (SALEE),

SALEE expansion mandrel, ROSA control computer, and standard sleeving system hydraulic expansion pressure unit. The Westinghouse sleeve delivery mandrel has an integral eddy current coil that senses the center of the TSP and enables the tool to automatically stroke into the install/expansion position. The sleeve delivery mandrel has been modified to properly position the center of the sleeve, and consequently the center of the expansion bladder, adjacent to the center of the TSP. The expansion process is computer controlled for consistency and repeatability. During the expansion process, the sleeve initially yields and contacts the tube. After the yielded sleeve contacts the tube, a volume-controlled expansion is applied in which a predetermined piston stroke is monitored for confirmation of the required volume. This process leads to comparable bulge sizes for both low and high yield tubing.

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Enclosure 1 PG&E Letter DCL-03-183 The expanded tubes will be inspected following application of the process to verify that proper expansion (proper bulge size) has been achieved. If the minimum acceptable bulge size has not been achieved either above or below the TSP, an additional tube must be selected for expansion. The under-expanded tube will provide added margin against TSP deflection during a postulated MSLB event. Due to the design of the expansion bladder, re-expansion of under-expanded joints is not feasible, since the increased sleeve to bladder gaps may cause bladder failure prior to complete expansion. Re-expansion should be attempted only if both expansions (above and below the TSP) are below the minimum acceptable value. This would be the case if a premature bladder failure occurred during the expansion process. For tubes where expansions are performed at more than one TSP, if it is necessary to select an additional tube for expansions due to an unacceptable bulge size, expansions are required only at the TSP intersection(s) having bulge sizes less than the acceptance limit. All expanded tubes will be plugged.

Post-expansion diameter verification of the expansions is required to ensure that the minimum stiffness requirements are met. A standard bobbin profilometry probe will be used to determine the mean diameter of the expansion maxima (above and below the TSP). The technique involves the use of a bobbin coil probe excited in differential and absolute modes at multiple frequencies, typically ranging from 10 kilo-Hertz (kHz) to 630 kHz. The lowest frequency penetrates outside of the sleeved tube and is used for SG landmark detection. The highest frequency has a very shallow depth of penetration and is used for the measurement of the diameter of the expansion. The bobbin probe integrates the signal response about the circumference of the tube and yields a mean diameter measurement at a given axial location.

A standard with expansions of known diameter is used to construct a calibration table that relates the diameter of the tube to the voltage of the eddy current response. The eddy current measurement of the inner diameter, on the average, meets the expected value within plus or minus 0.002 inch. This uncertainty on the bobbin profilometry results is acceptable and no adjustments are necessary to the bobbin data for field process applications. This shows that the tube ID can be reliably measured using eddy current methods. This measurement coupled with the knowledge of the strain experienced during the expansion process can be used to verify that the OD of the bulge falls within the desired process range.

Structural Integrity of Tube Expansion Joints After one cycle of operation, all TSP expansions at Braidwood-1 were inspected using the Plus Point coil and no indications were detected. The 27

Enclosure 1 PG&E Letter DCL-03-183 target expansion diameter at DCPP is less than the target expansion diameter used for Braidwood-1. Process improvements have been made to reduce the potential of axial misposition, which, in turn, determines bulge variance and the potential for large bulges. No cracking has been found in the hydraulic expansion at TSP intersections in the preheater region of Model E and D4 SGs in more than 10 years of plant operation.

Therefore, the likelihood of experiencing a circumferential crack in the parent tube at the TSP expansion is reduced for DCPP compared to Braidwood-1. Since no circumferential indications were detected in the TSP expansions at Braidwood-1 after one cycle, and smaller bulges will be made at DCPP, circumferential cracking is not considered an issue for up to three cycles of operation at DCPP. However, the DCPP TS will be modified to require in each locked SG, following two complete cycles of operation and if the SGs are not to be replaced at the outage, that 20 percent of the expanded tubes be Plus Point coil inspected at the expanded TSP locations. This will provide additional assurance that any circumferential indication in TSP expansions is detected.

After one operating cycle following TSP expansions at Braidwood-1, the top of the tubesheet (TTS) regions in expanded tubes were inspected using the Plus Point coil, and circumferential indications were detected at the top of the tubesheet region. The tube to tubesheet expansion process at Braidwood-1 was performed by hard rolling. The end of cycle (EOC) 6 inspection (first inspection after implementation of the 3-volt ARC) was the first use of the Plus Point probe at Braidwood-1. Prior TTS inspections at Braidwood-1 were performed with a rotating pancake coil probe. The results of the Braidwood-1 1997 inspection were discussed in a meeting between Commonwealth Edison (ComEd) and the NRC on April 29,1997.

ComEd concluded that the top of tubesheet circumferential indications in the expanded tubes were likely undetected indications from the prior inspection that had grown to Plus Point detectable levels at EOC 6. The signals of the circumferential indications in the expanded tubes did not represent a new degradation mechanism, but were, in fact, ODSCC at the roll transition. Subsequent evaluation indicated that the incidence of circumferential indications among the population of expanded tubes was independent of the number of expansions performed in a single tube.

Consequently, the detected circumferential indications at Braidwood-1 are considered to be independent of the expansion at TSP intersections.

The DCPP tubes were explosively expanded in the tubesheet by the Westinghouse explosive tube expansion (WEXTEX) process. The industry operating experience has shown that explosive expansions are significantly less susceptible to circumferential cracking than hardroll expansions. There are small numbers of PWSCC and ODSCC circumferential cracks detected in the DCPP SGs.

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Enclosure 1 PG&E Letter DCL-03-183 Historically in the WEXTEX region, there have been 16 circumferential ODSCC indications (in 16 tubes) and 11 circumferential PWSCC indications (in 10 tubes) through the 1R11 inspection for DCPP Unit 1.

One of the circumferential PWSCC indications is located below the W*

length and the tube is in service under the W* ARC. Thus, a total of 25 DCPP Unit 1 tubes have been plugged due to circumferential WEXTEX indications. There has been 1 circumferential ODSCC indication (1 tube) and 12 circumferential PWSCC indications (in 11 tubes) through the 2R1 1 inspection for DCPP Unit 2. Thus, a total of 12 Unit 2 tubes have been plugged due to circumferential WEXTEX indications. 40 circumferential indications were detected in 38 tubes, 35 of the tubes were located in the central region of the SG (approximately radius of row 30 tubes) and 3 of the tubes, with only PWSCC indications, were in more peripheral regions.

Results from the most recent DCPP inspections are used to define a detection threshold. At the DCPP 1R11 inspection, there were 4 circumferential ODSCC indications and I circumferential PWSCC indication in the WEXTEX region (W* length), of which the largest measured crack angle was 42 degrees. At the 2R11 inspection, there was 1 circumferential ODSCC indication in the WEXTEX region (W* length) with a measure crack angle of 78 degrees. The small Plus Point voltages (less than 0.35 volt) found for these indications implies that the cracks are both short and shallow.

The design of the DCPP TSP locking expansion was modified based on the Braidwood-1 operating experience. The objective was to reduce the residual stress in the tube due to expansion by reducing the required bulge diameter. To compensate for the expected loss of load carrying capability, a full wall thickness sleeve will be utilized for the DCPP process instead of the undercut sleeve utilized at Braidwood-1. The reduced expansion diameter reduces the residual stress in the tubes; thus the potential for circumferential cracking is reduced. The span lengths between the TTS and first TSP and between TSPs is longer in the Model 51 SGs at DCPP than in the Model D4 SGs at Braidwood-1. The longer span lengths would help to reduce the residual stress in the tube from the expansions. In addition, the large contact forces between the tube and packed TSP crevices are likely to reduce the effect of multiple expansions in the tube on the residual tensile stress at the TTS.

Although the likelihood of severing an expanded tube at the top of the tubesheet is negligible, a stabilizer will be installed at the top of the tubesheet of the expanded tubes to provide in-depth assurance that no new failure mode (damage to adjacent tubes from a severed tube) is introduced by the expansions. The top of the tubesheet stabilizer may be a sleeve or a cable. The sleeve incorporated in the TSP expansions also 29

Enclosure 1 PG&E Letter DCL-03-183 functions as a stabilizer in the unlikely event of a severed parent tube at an expanded tube location.

Axial cracking at the expansions has a low likelihood of occurrence, but even if it does occur, it would not significantly affect the bulge stiffness since TSP displacements must compress or attempt to flatten the bulge (i.e., extrude the bulge) for which the required forces would not be significantly influenced by an axial crack in the bulge. After expansion of a tube in the field, bobbin coil profilometry is used to confirm that acceptable expanded tube diameters have been achieved and that the expansions are properly located relative to the TSP.

SG Tube Structural Integrity Since the TSPs do not undergo any displacement relative to indications developed within the upper and lower planes of the TSPs during normal operation, tube burst at pressures less than three times the normal operating differential pressure is obviated by the presence of the TSP.

Therefore, the requirement of RG 1.121, for the resistance to burst relative to three times the normal operating differential pressure, (3AP), is inherently met. Thus the determination of the margin against tube burst only needs to be considered for the limiting accident conditions following a postulated MSLB event.

There are multiple modes of failure that can be classified as burst. An axial burst consists of a splitting of the tube parallel to the axis of the tube and is caused by hoop stresses due to internal pressure. An axial tensile failure, or tensile failure, consists of separating the tube into two parts through a cutting plane perpendicular to the tube axis (i.e., a guillotine break). The latter failure mode is caused by the unbalanced pressure load at the U-bend usually referred to as the end cap load.

The total POB of a tube with ODSCC at TSP elevations is the sum of the probability of axial burst plus the probability of tensile failure. The probability of tensile failure is negligible in SGs where the GL 95-05 ODSCC ARC voltage limits are applied because axial tensile failures are not expected based on plant operating experience. Only the probability of axial burst needs to be reported when the GL 95-05 voltage limits are used. Omission of consideration of the probability of tensile failure has also been approved by the NRC for application at South Texas Project Unit 2 when only a single cycle of operation with ODSCC ARC voltage limits above the GL 95-05 voltage limits was planned. The staff specifically noted in the safety evaluation for South Texas Project Unit 2 that the probability of axial burst and tensile failure should be combined for applications involving more than one cycle of operation.

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Enclosure 1 PG&E Letter DCL-03-183 The voltage repair limits for the current DCPP 2-volt ARC are based on the free span model as discussed in GL 95-05 and only axial burst is considered. For the current 2-volt ARC, no credit is taken for the TSPs reducing the likelihood of axial burst or tube leakage under transient and postulated accident conditions.

The voltage repair limits for the proposed 4-volt ARC are based on a locked TSP model in which credit is taken for the constraint provided by the TSPs under normal operation, transients, and postulated accident conditions, to reduce the likelihood of tube axial burst and leakage. The expansion of 30 tubes in each SG, at 196 intersections, prevents the 4 lowest TSPs from deflecting more than 0.15 inch in the hot-leg region of the SG tubes. This yields a POB of less than I 10-' for a single tube assumed to have through-wall crack length equal to, or greater than, the MSLB TSP displacement with the tip of the through-wall crack length assumed to be at the edge of the TSP. If all 13,552 intersections of the SG tube hot-legs with the 4 lowest TSPs had an indication, and each intersection displaced 0.15 inch, the total MSLB tube burst probability would still be less than 10-5. The methods to determine axial burst and tensile failure for the proposed 4-volt ARC are discussed separately in the following sections.

Deterministic Assessment of Axial Burst Failure The analyses for axial burst probability consist of consideration of the correlation of the burst pressure of through-wall cracks relative to crack length and the correlation of the burst pressure of ODSCC TSP indications relative to the nondestructive examination eddy current testing amplitude (i.e., the bobbin voltage). The concern for the potential of tube burst during a MSLB is based on the consideration that the pressure gradient in the SG will cause the TSPs to deform out of plane and expose TSP intersection tube ODSCC indications such that they behave as free-span indications without the constraint of the TSP. The evaluation of the likelihood of tube rupture due to axial burst (i.e., the POB), is based on the calculated deformations of the TSPs, the correlation of the burst pressure of tubes with free-span axial cracks to crack length, and the correlation of the burst pressure of tubes with free-span ODSCC indications to bobbin voltage.

The 4-volt ODSCC ARC will be applied to row and column locations where the TSP undergoes relatively minor displacements (i.e., less than 0.15 inch) during a MSLB initiated from either normal operating or hot standby conditions. Since the thickness of the TSP is 0.75 inch, it is unrealistic to treat each indication as though it would be fully exposed during a MSLB. For TSPs where the displacement is limited, the expected burst pressure can be calculated by considering a through-wall crack that 31

Enclosure 1 PG&E Letter DCL-03-183 is exposed by an amount equal to the deformation of the TSP. The POB for such a crack can be calculated using the correlation of the burst pressure to the crack length, adjusted as necessary to account for the constraint from the TSP. For larger deformations the POB is calculated as the larger of the values obtained by using both correlations developed for predicting the burst pressure (i.e., as a function of bobbin voltage or crack length). The rationale for this is that while the POB may increase significantly for longer through-wall cracks, the actual POB would be limited by that for a free-span ODSCC indication that can be predicted from the bobbin amplitude.

An analysis of burst test data for a variety of tube sizes indicates a strong correlation between burst pressure and through-wall crack length using an exponential relationship. This relationship was used to estimate the critical crack length of 0.89 inch for burst during a MSLB. The results imply that the limiting through-wall indication equal to the TSP thickness of 0.75 inch would not burst at MSLB conditions, and that the POB for an indication that extends only approximately 0.15 inch beyond the TSP surface is extremely low.

For tubes in which a portion of the length of the crack is restrained in the radial and circumferential directions, the burst pressure correlates with the exposed crack length. This is because the local condition for burst is the achievement of a critical opening of the crack at the crack tip. The critical crack tip opening displacement (CTOD) for through-wall cracks in thin walled tubing is on the order of the thickness of the tube, about 0.040 inch to 0.045 inch. In essence, the clearance between the OD of the tube and the ID of a TSP hole is not sufficient to permit the achievement of the critical CTOD for the end of the crack within the TSP at the pressures which would lead to the burst of cracks of significant length. Within the TSP, the crack would not be expected to extend in length beyond that associated with less-than-critical blunting of the crack tip. If the clearance between the inside of the TSP hole and the tube approaches zero, there can be no CTOD at that end of the crack and the strength of the cracked tube is slightly increased. If the clearance between the tube and the inside of the TSP is significant, the crack flanks may open and the crack would be expected to behave as though it were slightly longer than the exposed length. Finally, if the clearance is between the two extremes, the burst pressure may be slightly elevated or depressed depending on the value of the clearance.

In order to address the effect of the TSP on the burst pressure, a series of burst tests were performed to quantify the effect of the clearance between the tube and the TSP hole. Because an essential feature of the testing program was the presence of the TSP collar and the diametral clearance between the tube and the hole in the TSP, the tests were performed for 32

Enclosure 1 PG&E Letter DCL-03-183 the range of clearances that might be expected in Westinghouse fabricated SGs. The results from the tests to quantify the effect of a small clearance between the tube and the TSP hole verified that the burst pressure for a long crack, with a portion of the crack constrained by the TSP, would be similar to that of a free-span crack with a total length equal to the exposed length of the constrained crack. Therefore, the through-wall burst pressure correlation may be used to evaluate the POB of exposed cracks as a function of the length exposed if the diametral clearance between the tube and the hole in the TSP in the SG is small (i.e., on the order of 0.013 inch or less). The dented and packed TSP crevices at DCPP lead to the expectation that the actual diametral clearances are essentially zero.

The strengthening due to the TSP constraint would be significant for the larger clearances (greater than 0.013 inch), but not as significant as for the small clearance range. In other words, the burst pressure would be expected to be slightly less than that for free-span cracks with a total length equal to the exposed length of the constrained test specimens.

This has been confirmed by the results from the tests for larger gaps.

Over the range of exposures of the crack beyond the TSP of interest, the effect of the large clearance is to diminish the burst pressure for 0.875 inch diameter tubes. The data illustrate that no burst pressure adjustment is necessary for exposures of the crack beyond the TSP in the range of 0.1 inch to 0.2 inch and that a slight burst pressure adjustment is necessary for exposures of the crack beyond the TSP in the range of 0.2 inch to 0.3 inch. A burst pressure adjustment factor is used which bounds the data in the range of TSP displacement distance.

In summary, testing has shown that the tube burst pressure for an axial crack extending outside the TSP is a function of the exposed ODSCC crack length outside the TSP as opposed to the total ODSCC crack length. The testing also showed that the burst pressure for cracks that extend outside the TSP is only slightly reduced at larger than nominal diametral gaps between the tube OD and the inside of the TSP hole when compared to the burst pressure for a free-span crack of the same length.

The POB for a single axial ODSCC indication at a TSP intersection with an exposed crack length equal to the limited TSP displacement of 0.15 inch is negligible (less than 10-1°). Additional margin against burst of a single axial ODSCC indication during postulated accident conditions is by the SG tube inspections which ensure that all ODSCC indications remain in the portion of the tube confined to the TSP thickness.

Deterministic Assessment of Axial Tensile Failure The structural limit that determines tube repair limits for free-span indications is based on satisfying RG 1.121 margins for burst of a tube.

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Enclosure 1 PG&E Letter DCL-03-183 With the limited displacements of the lower TSPs, the constraint of the TSP reduces the tube burst probability to negligible levels and tube repair limits are not required to prevent axial tube burst. For the small TSP displacements and the expected tube degradation, the need for tube repair is dictated by the need to satisfy allowable MSLB leakage limits.

However, at some level of cellular or intergranular attack corrosion, it becomes possible for the axial loads resulting from the pressure differential across the tube to result in axial tensile severing of the tube.

This tensile load requirement establishes the applicable structural limit for tube expansion based on limited TSP displacement.

It is noted that the bobbin coil voltage is being used to assess the potential for a SG tube to fail circumferentially. An inherent assumption in the methodology is that there is a significant axial component in the SG tube flaw such that it can be detected with the bobbin coil because the bobbin coil is relatively insensitive to circumferentially oriented cracks in SG tubes. The degradation morphology that has been found in pulled tubes for ODSCC at TSP intersections is dominantly axial SCC with varying extent of cellular patches including the absence of cellular corrosion.

Cellular corrosion is a combination of axial and oblique angle cracks that form small cells of undegraded tubing within the crack pattern. The patterns of cellular corrosion have been characterized by radial grinds from the tube surface through the tube wall. The axial cracks are consistently deeper than the oblique cracks. That is, as the radial grinds progress through the wall, the crack pattern changes from cellular to multiple axial micro-cracks with the oblique cracks typically less than 50 percent to 60 percent of the tube wall when the axial cracks are near through-wall. The oblique cracks in the cellular pattern are typically less than 90 degrees (i.e., not circumferential) and the bobbin coil voltage responds to both the axial and oblique cracks.

The bobbin voltage responds primarily to the deepest and longest axial cracks. In correlating bobbin voltage with the axial tensile capability of the tube, it is assumed that the cellular pattern will increase in circumferential extent in some close proportion to the length and depth of the dominant axial cracks. Since this is not generally the case, it can be expected that the spread in the correlation (range of tensile force capability at a given voltage) will be significant and greater than that found for the axial burst pressure correlation with voltage. The results of calculations indicate the overall structural limit for axial separation at 3AP is well in excess of 100 volts which provides a safety factor of 25 to circumferential rupture for indications with an amplitude of less than or equal to 4 volts. Also, all indications with an amplitude of less than or equal to 4 volts would be expected to have significant margin to burst relative to the RG 1.121 3AP limit during normal operation.

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Enclosure 1 PG&E Letter DCL-03-183 Probabilistic Structural Integrity Assessments Probabilistic analyses for the potential for SG tube burst following a MSLB have been performed. The probabilistic analyses include assessment of axial burst failure and axial tensile failure, which are discussed below.

Probabilistic Assessment of Axial Burst Failure The method to calculate the burst probability with TSP displacements is contained in Section 8.5 of WCAP-16170-P, Revision 0. The burst pressure of a through-wall indication extending outside the TSP has a burst pressure almost equal to the burst pressure for a free span crack at the length extending outside the TSP. Based on this, the burst probability for TSP displacements with postulated through-wall cracks can be calculated as that associated with a crack length equal to the TSP displacement. A very conservative assumption was made that all intersections at all hot-legs of the lower 4 TSPs have a through-wall crack length at least equal to the TSP displacement. This assumption is applied to determine the allowable TSP displacement so that the proposed 4-volt ARC envelops all possible tube degradation. There are 13,552 total intersections for the lower 4 hot-leg TSPs with 3,388 tube intersections at each TSP. For simplicity, because the individual burst probabilities are so small, the burst probabilities are calculated as the number of TSP intersections ) times the single tube burst probability for a through-wall crack equal to the maximum TSP displacement (i.e., less than 0.15 inch).

The maximum TSP displacement is limited by the expanded SG tubes.

The burst probability contribution from all intersections at all hot-legs of the lowest 4 TSPs would be less than 10-5 even under the extreme assumption of exposed through-wall cracks at the TSP displacement design limit for every TSP intersection. Therefore, the burst probability contribution from all intersections at all hot-legs of the lowest 4 TSPs can be ignored in the burst probability analysis.

As stated in GL 95-05, the ARC for ODSCC indications is only applicable to cracks that are fully confined within in the TSPs. GL 95-05 was based, in part, on experience that ODSCC on tubes inside TSPs had not produced cracks with significant extensions beyond the confines of the TSPs into the free-span portion of the tube. Based on past inspection experience, significant extension of a crack beyond the confines of the TSP is not expected. However, if ODSCC indications are identified which extend beyond the confines of the TSPs, the NRC will be notified prior to returning the SGs to service as required byTS 5.6.10.d.4 under the current 2-volt ODSCC ARC. Section 2.2.8 of WCAP-1 6170-P, Revision 0, requires that if one or more ODSCC indications at TSP intersections are found in the NDE inspection to extend beyond the edge of the TSP, the 35

Enclosure 1 PG&E Letter DCL-03-183 indications shall be evaluated for significance to safety and risk, and the results reported to the NRC prior to restart. The reporting of the results of the evaluation is considered to be a part of the NRC notification, and therefore the current TS 5.6.10.d.4 reporting requirement is considered to be sufficient to address the potential for significant extension of a crack beyond the confines of the TSP for the proposed 4-volt ODSCC ARC.

Probabilistic Assessment of Axial Tensile Failure The results of an evaluation of the probability of rupture as a function of indication amplitude for a 7/8-inch OD tube is provided in Figure 8-6 of WCAP-16170-P, Revision 0. The evaluation method is similar to that used previously for 3/4-inch OD tubes as documented in EPRI report NP-7480-L, Addendum 2, "Steam Generator Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates Database for Alternate Repair Limits, 1998 Database Update," dated April 1998, except that the larger tube area and differences in material properties were considered.

The probability of rupture of a single indication with a bobbin amplitude of 30 volts at a MSLB differential pressure of 2,405 pounds per square inch is 5xI0l 7per Figure 8-6 of WCAP-16170-P, Revision 0, and the overall structural limit for axial separation at 3 times normal operating differential pressure is well in excess of 100 volts. Also, all indications with a bobbin amplitude less than or equal to 4 volts would be expected to have significant margin to burst relative to the RG 1.121 3 times normal operating differential pressure limit.

To date, 5 tubes have been removed from DCPP Units 1 and 2 for destructive examination and to satisfy SG ARC applications. None of these tubes have had significant cellular corrosion including the recent 5.09 and 21.5 bobbin volt indications removed during 2R1 1.

Consequently, tensile tearing tests have not been performed for the DCPP pulled tubes.

Although cellular corrosion does not appear to be present to a significant extent at DCPP, the burst probability for axial tensile tearing will be calculated for the intersections of the SG tube hot-legs and the lower 4 TSPs in which the 4-volt ARC is applied. This conservatively assumes that cellular corrosion will occur at DCPP with the 4-volt repair limit. Upon implementation of the 4-volt ARC, the axial tensile tearing rupture probability for the intersections of the SG tube hot-legs and the lower 4 TSPs will be combined with the freespan burst probability for axial ODSCC indications for the intersections of the SG tube hot-legs and the top three TSPs and the intersections of the SG tube cold-legs and the seven TSPs. This requirement will be included in a new TS 5.6.10.d.5 reporting requirement which will state "For locked SGs, the burst probability is calculated using the methodology of WCAP-16170-P, 36

Enclosure 1 PG&E Letter DCL-03-183 Revision 0, that is, the axial tensile tearing rupture probability at tube support plate intersections at the 4 lowest hot-leg tube support plates is combined with the conditional burst probability at all other tube support plate intersections."

Leak Integrity Assessment Similar to the current 2-volt ARC, with use of a 4-volt ARC SG tubes with through-wall or near through-wall cracks may remain in-service creating potential primary-to-secondary tube leakage during normal operating or postulated postaccident conditions. TS 3.4.13, RCS Operational Leakage," limits the normal SG primary-to-secondary leak rate limit to 150 gallons per day (gpd) through any one SG. The limit for accident induced leakage through SG tubes is based on the FSARU Section 15.5.18.1 MSLB radiological consequences analysis. The current limit on postaccident induced leakage is 10.5 gallons per minute.

Leakage analyses to predict potential accident induced leakage following MSLB accident conditions have been performed. Normal operating leakage and the analyses of accident induced leakage are discussed below.

Normal Operational Leakage TS 3.4.13 limits the normal SG primary-to-secondary leak rate limit to 150 gpd through any one SG. This TS limits normal operating tube leakage and provides assurance that a single crack leaking 150 gpd would not propagate to a SG tube rupture under the stress conditions of a LOCA or a main steam line rupture. The primary-to-secondary leak rate limit combined with an effective leak rate monitoring program provides additional assurance that should a significant leak be experienced in-service, it will be detected, and the plant shut down in a timely manner.

Leakage Under Accident Conditions The leakage under postaccident conditions includes leakage from degraded tubes at locations that are not in locked TSPs and leakage at locations that are in locked TSPs for which the 4-volt ARC is applied. A methodology has been developed to combine the leakage from locations that are not in locked TSPs with the leakage from locations that are in locked TSPs based on adding the individual ninety-fifth percentile results for each location. The methodology is described in Section 8.0 of WCAP-1 6170-P, Revision 0, and is summarized below.

The potential leakage from degraded tubes at locations that are not locked TSPs include cracks, axial burst, and axial tensile failure of the degraded 37

Enclosure 1 PG&E Letter DCL-03-183 tube. There are no IRBs associated with TSPs that are not locked; thus, IRB leakage is not considered at locations that are not in locked TSPs.

The leakage due to axial burst and axial tensile failure has been excluded in the past since the total POB or tensile failure events have been extremely small. For TSPs that are not locked, the probability of a tensile failure is almost nonexistent and thus leakage due to tensile failure has no meaningful contribution to leak rate and is not considered. For TSPs that are not locked, the probability of axial burst remains small. Since the potential leak rate from a single burst is relatively large (e.g., greater than 100 gpm), a value of 100 gpm will be used as the leak rate for indications predicted to burst following a postulated MSLB accident. The specific value of the leak rate for indications predicted to burst (e.g., 100 gpm) is not meaningful to the prediction of the ninety-fifth percentile leak rate value as long as the number of affected tubes is small. Since the total POB is limited to 0.01 at 95 percent confidence, the actual number of SG Monte Carlo simulations in which tube burst is predicted to occur is limited. The leakage methodology at locations that are not in locked TSPs will be based on GL 95-05 methodology for cracks as modified for the leak rate for indications predicted to burst.

The potential leakage from degraded tubes at locations that are locked TSPs include cracks, axial burst, IRBs, and axial tensile failure of the degraded tube. For TSPs which are locked, the probability of axial burst is almost nonexistent due to the constraint provided by the TSPs; thus, leakage due to axial burst has no meaningful contribution to leak rate and is not considered. For TSPs which are locked, the probability of tensile failure remains small. Since the potential leak rate from a single burst is relatively large (e.g., greater than 100 gpm), a value of 100 gpm will be used as the leak rate for indications predicted to have a tensile failure following a postulated MSLB accident.

Leakage can occur from an indication in a locked TSP due to an overpressurized tube that expands within the TSP until the crack flanks contact the ID of the TSP drilled hole. The term "overpressure" refers to the fact that burst would be expected if the indication were located in the free-span portion of the tube. With the use of increased voltage repair limits associated with the 4-volt ARC, the probability of occurrence of indications, which could experience increased crack flank deflections, referred to as overpressurization, is increased. Therefore, the leak rate model needs to account for the leak rate from potentially overpressurized tubes. The leak rate model for IRBs in overpressurized tubes is based on a modification to the ODSCC ARC leakage model approved for the 2-volt ARC in Amendment Nos. 124 and 122 to Facility Operating License Nos. DPR-80 and DPR-82, respectively, for DCPP Units 1 and 2 in a letter to PG&E dated March 12, 1998. The method to determine the leakage for IRBs involves a Monte Carlo simulation of the distribution of indications, 38

Enclosure 1 PG&E Letter DCL-03-183 the uncertainties in the measurement of the indications, and the growth for the indications. The Monte Carlo simulation includes the following steps:

1. Determine the end-of-cycle voltage distribution for the ODSCC indications in accordance with GL 95-05.

2. Determine the free-span burst pressure for each of the ODSCC indications.

3. If the simulation indication burst pressure is less than the differential pressure during a postulated MSLB, the probability of leakage of the indication is assumed to be 1 and the indication leak rate is assumed to be 5.0 gpm. The leak rate of 5.0 gpm based on tests is conservatively applied independent of the indication size.

4. If the indication is predicted not to burst, the probability of leakage is determined from the logistic correlation of the probability of leakage to the common logarithm of the bobbin amplitude, a random uniform deviate is generated to determine if the indication leaks, and if the indication leaks, the leak rate is calculated from the correlation leak rate versus bobbin amplitude.

5. Steps 2 to 4 are repeated for each ODSCC indication in the SG locked TSPs and the leakage due to all indications is summed to determine the total SG leak rate.

6. Steps 2 to 5 are repeated in a Monte Carlo simulation to determine the total SG leak rate distribution.

7. The total leak rate is calculated on a 95 percent confidence bound on 95 percent of the possible population of leak rates based on results of greater than 100,000 Monte Carlo simulations of the total leak rate.

The total leak rate for each SG is determined from the sum of the leak rate due to cracks and axial burst, for hot-leg and cold-leg indications at TSPs which are not locked which are left in service under the current 2-volt ARC, and of the leak rate due to cracks, IRBs, and tensile failure, for hot-leg indications at TSPs which are locked which are left in service under the proposed 4-volt ARC. The leak rates will be applied such that the total leak rate includes the joint probability of all the potential failure modes. The ninety-fifth percentile total leak rate at a 95 percent confidence bound is determined using an ordered array of total leak rates from all of the simulations of the SG. The total leak rate is compared to 39

Enclosure 1 PG&E Letter DCL-03-183 the MSLB radiological consequences analysis limit on postaccident induced leakage, currently 10.5 gpm.

In summary, a change is made to the current GL 95-05 leak rate methodology for the current 2-volt ARC to apply a large leak rate of 100 gpm for indications predicted to burst as free-span indications at TSP elevations which are not locked. The changes to the GL 95-05 methodology for the proposed 4-volt ARC are to apply a bounding leak rate of 5 gpm for IRBs and to apply a large leak rate of 100 gpm for hot-leg indications that experience a tensile failure at locked TSP elevations. The total leak rate, which considers all potential failure modes, will be applied for ODSCC indications.

To implement the revised leakage requirements to support the changes to the GL 95-05 leak rate methodology for the current 2-volt ARC and the new 4-volt ARC, a new TS Section 5.6.10.d.1 is added which states; "For locked SGs, the estimated leakage from tube support plate intersections at the 4 lowest hot-leg tube support plates shall be calculated using the methodology of WCAP-16170-P, Revision 0, and combined with leakage from all other tube support plate intersections as calculated using the methodology of Generic Letter 95-05 modified to include a large leak rate assigned to indications predicted to burst in a Monte Carlo simulation as described in WCAP-16170-P, Revision 0." The new sentence is added to define the method to be used to calculate the leak rate for TSPs which are locked and for TSPs which are not locked. The NRC will be notified prior to returning the SGs to service per current TS 5.6.10.d.1 if the estimated leakage based on the projected end-of-cycle exceeds the leak limit determined from the licensing basis dose calculation for the postulated MSLB. This is consistent with the reporting requirements of GL 95-05.

Tube Leakage Tests The test data to support the 5 gpm leak rate used for IRBs in locked TSPs is discussed in Section 7.0 of WCAP-16170-P, Revision 0. The tests for IRBs are documented in EPRI Report TR-107625, "Steam Generator Indications Restricted from Burst (IRB) Leak Rate Tests," (Draft) April 1998 and were performed for both 3/4-inch OD and 7/8-inch OD tubes.

The limiting SG pressure differential for DCPP is 2405 pounds per square inch differential (psid), based on the PORV setpoints and uncertainties.

Based on the IRB tests, the bounding leak rate for 2,405 psid is 5.0 gpm, including the effects of a TSP offset of 0.21 inch.

The IRB tests considered TSP offsets of up to 0.21 inch, including through-wall offsets up to 0.17 inch, which conservatively bounds the maximum TSP displacement of 0.15 inch for DCPP to support the 4-volt ARC. The measured leak rates for the tests do not depend on tube size, 40

Enclosure 1 PG&E Letter DCL-03-183 and thus the combined test data for both 3/4-inch OD and 7/8-inch OD tubes can be applied equally for both tube sizes. No adjustment to the bounding leak rate is required for potential multiple indications since leakage from a tube with two cracks is dominated by the crack with the longest through-wall length and the probability of multiple through-wall cracks occurring at the location of the maximum TSP offset is extremely low.

The bounding leak rate is conservative due to the large tube/TSP gap used in the tests, which was a 95 percent confidence bound on the expected SG tube/TSP gap, and due to the orientation of the cracks within the TSP with the full gap opposite the crack to maximize the crack opening area. For DCPP, the tube to TSP gaps are essentially zero at operating temperatures due to dented and packed crevices. The DCPP crevice conditions would prevent crack opening and limit leakage to negligible levels compared to free-span leak rates and the bounding IRB leak rate of 5 gpm.

In summary, based on an evaluation of the IRB tests and analysis uncertainties, the bounding IRB leak rate of 5 gpm is conservatively high by 7 to 10 percent. In addition, for the calculation of IRB leak rate, the leak rate of 5.0 gpm will be conservatively applied independent of the indication size. These conservatisms will ensure the predicted overall IRB leak rate is conservative.

4.6 Summarv Adequate structural and leakage integrity can be provided for degraded tubes left in service under the proposed 4-volt ARC until the DCPP SGs are replaced. The estimated leakage and burst probability based on the projected end-of-cycle voltage distribution due to all ARC, including the proposed 4-volt ARC, will be determined each cycle and reported to the NRC in the SG 90-day report. The NRC will be notified prior to returning the SGs to service if the total leakage due to all ARC exceeds the leak limit determined from the licensing basis dose calculation for the postulated MSLB or if the total burst probability due to all ARC exceeds 0.01. In addition, the NRC will be notified prior to returning the SGs to service if indications left in service under the 4-volt ARC are identified that extend beyond the confines of the TSP, if a significant increase in TSP ligament cracking in expanded tubes is found, or if circumferential indications are found in expanded tubes.

The proposed 4-volt ARC includes TS requirements for Plus Point coil inspection of 100 percent of the expanded tubes at the top of the tubesheet following the expansion process. If the SGs are not to be replaced at the outage following two complete cycles of operation, 41

Enclosure 1 PG&E Letter DCL-03-183 20 percent of the expanded tubes shall be Plus Point coil inspected at the top of the tubesheet and expanded TSP locations, and the inspection shall be expanded to 100 percent of the expanded tubes in any SG that a circumferential indication greater than 100 degrees is found at the top of the tubesheet or above or below the tube support plate edges at tube support plate locations. These requirements ensure the long term integrity of the expanded SG tubes which support the 4-volt ARC.

The implementation of the 4-volt ARC includes a SG internals inspection plan to assure the load path assumed in the TSP thermal hydraulics remains intact. The inspection plan includes inspections for upcoming refueling outages, inspections to be performed during 1R12 in one SG, inspections to be performed in all SGs during the outage in which TSP locking is implemented, and inspections to be performed every outage following TSP locking until the SGs are replaced. Based on the results of previous SG internals inspections at DCPP, this inspection plan is considered to be sufficient to ensure long term integrity of the SG internal components and TSPs and to support a 4-volt ARC until SG replacement.

The proposed change includes the application of LBB to the MSL piping inside containment. An analysis, performed per the requirements of GDC 4 of Appendix A to 10 CFR 50 and the limitations and acceptance criteria for LBB contained in NUREG-1061, Volume 3, demonstrates the probability of pipe rupture is extremely low under conditions consistent with the design basis of the piping. In addition, the DCPP leak detection system is capable of detecting a leak rate that is 10 times lower than that due to the leakage flaw size in the MSL piping determined by the LBB analysis. Therefore, large MSLBs may be excluded from consideration for determining the loads on the SG TSPs following a MSLB. LBB will only be applied to the main steam piping for determination of the loads on the SG TSPs to support the 4-volt ARC and is only needed until the DCPP SGs are replaced.

5.0 REGULATORY ANALYSIS

5.1 No SiQnificant Hazards Consideration PG&E has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

42

Enclosure 1 PG&E Letter DCL-03-183

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

A 4-volt steam generator (SG) bobbin coil probe voltage-based alternate repair criteria (ARC) for axial outside diameter stress corrosion cracking (ODSCC) at tube support plate (TSP) locations is proposed for the hot-leg region of the SG tube at the 4 lowest TSP locations (TSPs 1 through 4).

In order to implement the proposed 4-volt ARC, sufficient SG tubes will be expanded in the hot-leg region of TSPs 1 through 4 to limit the TSP deflections following a limiting main steam line break (MSLB) event.

SG tubes pass through holes drilled in the TSP. The inside diameter of the drilled holes closely approximates the outside diameter of the tubes.

Generally, the TSP precludes those tube spans within the drilled holes from deforming beyond the diameters of the drilled holes, thus, precluding tube burst in the restrained regions. However, design basis MSLB events may vertically displace a TSP, removing its support from the tube spans passing through it. For TSPs at hot-leg locations in which sufficient SG tubes have been expanded at the TSP intersection, the deflections of the TSP following a limiting MSLB event are small, the TSPs remain essentially stationary during all conditions, and the SG tube spans within the drilled TSP holes are restrained. Thus, for intersections of SG tube hot-legs and TSPs I through 4, axial tube burst is eliminated as a credible event and the larger bobbin voltage for the, proposed 4-volt ARC can be allowed while still meeting the tube structural requirements of Regulatory Guide (RG) 1.121.

For the calculated displacement of the affected TSPs following a limiting design basis MSLB event, based on application of leak-before-break to the main steam system piping inside containment, tube hot-leg spans enclosed within TSPs 1 through 4 have a tube burst probability of much less than 10 5 collectively. This is orders of magnitude less than the 10-2 probability-of-burst criterion specified by Generic Letter (GL) 95-05 and represents negligible axial tube burst probabilities for affected tube hot-leg spans intersecting TSPs. Thus, repair limits to preclude burst are not needed and tube repair limits for intersections of SG tube hot-legs and TSPs 1 through 4 may be based primarily on limiting leakage to acceptable levels during accident conditions.

Cracks that include cellular corrosion may yield under axial loads, resulting in tensile tearing of the tube at that location. A tensile load requirement to prevent this establishes a structural limit for the tube expansion based ODSCC ARC. In order to establish a lower bound for the structural limit, tensile tests were used to measure the force required 43

Enclosure 1 PG&E Letter DCL-03-183 to separate a tube that exhibits cellular corrosion. Additionally, pulled SG tubes with cellular and/or inter-granular attack (IGA) tube wall degradation were evaluated and the tensile strength of the tube was conservatively calculated from the remaining noncorroded cross-section of the tube. The tensile strength calculation assumed that the degraded portions do not contribute to the axial load carrying ability of the tube. Data from these tests shows that circumferential cracks exhibiting bobbin coil probe indication voltages greater than 100 volts at the lower 95 percent confidence level require tube pressure differentials above the operating limit of 3-times normal operating differential pressure in order to produce circumferential ruptures (i.e., axial separation at the plane of the crack due to axial tensile tearing). The proposed 4-volt ARC has a safety factor of 25 to circumferential ruptures, which ensures the 4-volt ARC does not significantly increase the chances of a steam generator tube rupture (SGTR) at intersections of SG tube hot-legs and TSPs I through 4.

In addressing the potential combined Loss-of-coolant accident (LOCA) and earthquake effects on SG components as required by General Design Criterion (GDC) 2 of Appendix A to 10 CFR 50, analysis has shown that SG tube deformation or collapse may occur in certain regions of the SG.

SG tube collapse reduces RCS flow and could cause partial through-wall tube cracks to become full through-wall tube-cracks during tube deformation or collapse resulting in potential secondary-to-primary in-leakage to the reactor coolant system (RCS). Tubes for which deformation may occur are excluded from application of the voltage-based ARC per current TS 5.5.9.d.1 .j (iv). TS 5.5.9.d.1 j (iv) will continue to apply and is not adversely affected by the 4-volt ARC. Therefore tubes for which deformation may occur will not be left in service under the 4-volt ARC.

GL 95-05 states that licensees must perform SG tube postaccident leak rate and SG tube burst probability analyses before returning to power from outages during which they perform SG inspections. Licensees must include the results in a report to the NRC within 90 days after restart. If an analysis reveals that postaccident leak-rate or burst-probability exceeds limits, the licensee must report it to the NRC and assess the safety significance of this finding.

For the proposed 4-volt ARC, the axial tensile tearing tube rupture probability is calculated for indications found at intersections of tube hot-legs and TSPs I through 4. The sum of MSLB axial tube burst probability for cold-leg TSP intersections, MSLB axial tube burst probability for hot-leg intersections at TSPs 5 through 7, axial tensile tearing tube rupture probability for TSPs 1 through 4, and the burst probability for indications left in service under other ARCs must be compared to the GL 95-05 reporting value of 10-2. Due to the negligible 44

Enclosure 1 PG&E Letter DCL-03-183 burst probability for axial ODSCC indications at intersections of tube hot-legs and TSPs 1 through 4, calculation of the axial burst probability is not required for these indications.

The design basis MSLB outside of containment produces the limiting radiological consequence from any SG tube leakage due to SG tube indications that are postulated to exist at the initiation of an accident.

Verification prior to each operating cycle, that the sum of MSLB leak rates from indications left in service under all ARC (including the proposed 4-volt ARC) are less than the leak rate limit assumed in the MSLB radiological consequences analysis, will ensure that site boundary doses for this accident remain within an acceptable fraction of the guidelines of Title 10 of the Code of Federal Regulations, Part 100, (10 CFR 100) and that doses to the control room operators remain within the 10 CFR 50 Appendix A GDC 19 limits.

The application of leak-before-break (LBB) to the MSL piping inside containment does not alter the way in which plant equipment is operated and cannot initiate an accident. The application of LBB to the main steam system does not affect the plant operating conditions and will not challenge the ability of the main steam system to perform its design function or to mitigate an accident.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different accident from any accident previously evaluated?

Response: No.

Use of the proposed SG tube 4-volt ARC does not significantly change the operating conditions of the SG. Application of the 4-volt ARC does not significantly increase the probability of either single or multiple tube ruptures. SG tube integrity remains adequate for all plant operating conditions. The GL 95-05 SG tube integrity limits will be confirmed through in-service inspection and monitoring of primary-to-secondary leakage.

The Diablo Canyon Units 1 and 2 Technical Specifications (TS) impose a normal SG primary-to-secondary leak rate limit of 150 gallons per day (gpd) per SG to minimize the potential for excessive leakage during all plant conditions. The 150 gpd limit provides added margin to accommodate contingent leakage should a stress corrosion crack grow at a greater than expected rate or extend outside the TSP. The proposed 4-volt ARC does not adversely impact the TS 150 gpd limit. Normal 45

Enclosure 1 PG&E Letter DCL-03-183 operating leakage is not expected to significantly increase due to indications left in service under the proposed 4-volt ARC.

The application of LBB to the MSL piping inside containment does not involve any physical alteration to the plant or any change in which the plant is operated which could introduce a new failure mode. The use of LBB does not involve plant equipment being operated in a different manner.

Therefore, the proposed change does not create the possibility of a new or different accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

RG 1.121 describes a method for meeting GDCs 14, 15, 31, and 32 of Appendix A to 10 CFR 50 by reducing the probability or consequences of SGTR through application of criteria for removing degraded tubes from service. These criteria set limits of degradation for SG tubing through in-service inspection. Analyses show that tube integrity will continue to meet the criteria of Regulatory Guide 1.121 after implementation of the proposed 4-volt ARC. Even under the worst case ODSCC occurrence at TSP elevations left in service under the 4-volt ARC, the 4-volt ARC will not cause or significantly increase the probability of a SGTR event.

Verification prior to each operating cycle, that the sum of MSLB leak rates from indications left in service under all ARC (including the proposed 4-volt ARC) are less than the leak rate limit assumed in the MSLB radiological consequences analysis, will ensure that site boundary doses for this accident remain within an acceptable fraction of the guidelines of 10 CFR 100 and that doses to the control room operators remain within the limits of GDC 19 of Appendix A to 10 CFR 50.

Inspections conducted for the proposed 4-volt ARC are the same as required by GL 95-05 with adjustment of the rotating pancake coil inspection requirements for hot-leg TSPs 1 through 4 intersections to reflect the higher 4-volt ARC limit. All hot-leg TSPs 1 through 4 intersections with bobbin coil voltages greater than 4 volts will be inspected with Plus Point coil and a Plus Point coil minimum sample inspection of intersections with bobbin indications less than or equal to 4 volts will be applied to hot-leg TSPs 1 through 4. The Plus Point coil data will be evaluated to confirm that the principal degradation mechanism continues to be ODSCC.

46

Enclosure I PG&E Letter DCL-03-183 Plugging SG tubes reduces RCS flow margin. The 4-volt ARC will reduce the number of tubes that must be plugged. Thus, the 4-volt ARC will conserve RCS flow margin, preserving operational and safety benefits that would otherwise be reduced by unnecessary plugging.

The application of LBB to the MSL piping inside containment will not adversely affect operation of plant equipment and will not result in a change to design basis accident initial conditions or the setpoints at which protective actions are initiated. With application of LBB, the main steam system will continue to perform its function as assumed in the accident analyses.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, PG&E concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of 'no significant hazards consideration" is justified.

5.2 Applicable Requlatorv Requirements/Criteria GDC 14 of Appendix A to 10 CFR 50, "Reactor Coolant Pressure Boundary," contains requirements applicable to SG tubes since they are part of the reactor coolant pressure boundary. GDC 14 requires that the reactor coolant pressure boundary be designed, fabricated, erected, and tested in order to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross failure.

RG 1.121 provides guidance for determining the minimum wall thickness at which a SG tube should be plugged. The RG 1.121 performance criteria recommend that the margin of safety against SGTR under normal operating conditions should not be less than 3 at any tube location where defects have been detected. The margin of safety against tube failures under postulated accident conditions should be consistent with the margin of safety determined by the stress limits specified in Section II of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. Analyses have been performed to ensure the RG 1.121 requirements continue to be met for the 4-volt ARC.

GL 95-05 provides the requirements for voltage-based ARC for predominately axial ODSCC confined to the thickness of the TSPs.

GL 95-05 sets voltage-based repair limits based on empirically derived correlations between a nondestructive inspection parameter, the bobbin coil voltage, and tube burst pressure and leak rate. The GL 95-05 voltage-based ARC allows a SG tube with an ODSCC indication, which 47

Enclosure 1 PG&E Letter DCL-03-183 has a bobbin coil voltage which is lower than the repair limit to remain in service. The GL 95-05 guidance ensures that tube structural and leakage integrity continue to be maintained at acceptable levels consistent with the requirements of 10 CFR 50 and 10 CFR 100. The 4-volt ARC will continue to meet the structural and leakage integrity limits required by GL 95-05. The POB and normal and postaccident leakage will be calculated for all indications left in service under all ARC, including the 4-volt ARC, to ensure the GL 95-05 and 10 CFR 100 limits are met. The indications left in service under the 4-volt ARC will be restricted from burst by the expansion of certain SG tubes into TSPs 1 through 4.

GDC 4 of Appendix A to 10 CFR 50 allows the use of fracture mechanics methodology to exclude from structural design consideration the dynamic effects of pipe ruptures in nuclear power plants, provided it is demonstrated that the probability of pipe rupture is extremely low under conditions consistent with the design basis of the piping. The demonstration of low probability of pipe rupture utilizes a deterministic fracture mechanics analysis that evaluates the stability of postulated, small, through-wall flaws in piping and the ability to detect leakage through the flaws before the flaw could grow to unstable sizes and break the pipe.

This concept is referred to as LBB. The limitations and acceptance criteria for LBB are contained in NUREG-1 061, Volume 3. The fracture mechanics analysis for the proposed application of LBB to the MSL piping inside containment, to limit the maximum MSLB induced forces on the SG TSPs, meets the requirements of GDC 4 of Appendix A to 10 CFR 50 and NUREG-1061, Volume 3. These requirements include a margin of 10 on leakage detection capability, a margin of square root of 2 for loads, a factor of 2 between the leakage-size flaw postulated for normal loads and the critical size flaw postulated for normal plus safe shutdown earthquake loads, and redundant leakage detection capability.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

PG&E has evaluated the proposed amendment and has determined that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 48

Enclosure 1 PG&E Letter DCL-03-183 10-CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

7.1 References

1. Westinghouse Electric LLC WCAP-16170-P, Revision 0, "Diablo Canyon SG Alternate Repair Criteria Based On Limited Tube Support Plate Displacement," dated October 2003 (proprietary).

2. Westinghouse Electric LLC WCAP-1 6170-NP, Revision 0, "Diablo Canyon SG Alternate Repair Criteria Based On Limited Tube Support Plate Displacement," dated October 2003 (nonproprietary).

3. Structural Integrity Associates, Inc. Report No. SIR-03-146, Revision 1, "Leak-Before-Break Evaluation Main Steam Piping Inside Containment Diablo Canyon Power Plant Units 1 and 2," dated December 2003.

4. Regulatory Guide 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," dated August, 1976.

5. Title 10 to the Code of Federal Regulations, Part 50, (10 CFR 50),

October 25, 2000.

6. Generic Letter 95-05, Voltage-Based Repair Criteria For Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking," dated August 3, 1995.

7. General Design Criteria 4 of Appendix A to Part 50 to Title 10 to the Code of Federal Regulations, October 27, 1987.

8. NUREG-0800, Standard Review Plan, Section 3.6.3, "Leak-Before-Break evaluation Procedures," dated March 1987.

9. Westinghouse Electric LLC WCAP-14277, Revision 1, "SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections," dated December 1996.

10. Amendment Nos. 124 and 122 to Facility Operating License Nos.

DPR-80 and DPR-82, respectively, "Issuance of Amendments for Diablo Canyon Nuclear Power Plant, Unit No. I (TAC No. M97254) and Unit No. 2 (TAC No. M97255)," dated March 12,1998.

11. PG&E letter DCL-97-034, "License Amendment Request 97-03, Voltage-Based Alternate Steam Generator Tube Repair Limit for Outside Diameter Stress Corrosion Cracking at Tube Support Plate Intersections," dated February 26,1997.

49

Enclosure 1 PG&E Letter DCL-03-183

12. NUREG-1061, Volumes 1-5, "Report of the U.S. Nuclear Regulatory Commission Piping Review Committee," dated April 1985.

13. Westinghouse Electric LLC WCAP-1 5163, Revision 1, uSouth Texas Project Unit 2; 3V Alternate Repair Criteria Application of Bounding Analysis and Tube Expansions," dated January 2001.

14. License Amendment No. 114 to Facility Operating License No. NPF-80, "South Texas Project (STP) Unit 2 - Issuance of Amendment Revising the Technical Specifications to Implement 3-volt Alternate Repair Criteria for Steam Generator Tube Repair (TAC No. MA8271)," dated March 8, 2001.

15. Westinghouse Electric LLC WCAP-14273, "Technical Support for Alternate Plugging Criteria with Tube Expansion at Tube Support Plate Intersections for Braidwood-1 and Byron-1 Model D4 Steam Generators," dated February 1995.

16. License Amendment No. 69 to Facility Operating License No. NPF-72 for Braidwood Nuclear Power Station Unit 1 and License Amendment No. 77 to Facility Operating License No. NPF-37 for the Byron Nuclear Power Station Unit 1, "Issuance of Amendments (TAC Nos. M96498, M96499, M96500 and M96501)," dated November 9, 1995

17. License Amendment No. 82 to Facility Operating License No. NPF-72 for Braidwood Nuclear Power Station Unit 1, "Issuance of Amendments (TAC Nos. M91671, M91672, M91673 and M91674)," dated May 14, 1997.

18. NRC Information Notice 96-09, "Damage in Foreign Steam Generator Internals," dated February 12, 1996.

19. NRC Information Notice 96-09, Supplement 1, 'Damage in Foreign Steam Generator Internals," dated July 10, 1996.

20. Generic Letter 97-06, "Degradation of Steam Generator Internals,"

dated December 30, 1997.

21. EPRI report NP-7480-L, Addendum 2, "Steam Generator Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates Database for Alternate Repair Limits, 1998 Database Update," dated April 1998.

22. EPRI report TR-1 07625, uSteam Generator Indications Restricted from Burst (IRB) Leak Rate Tests," (Draft), dated April 1998.

23. NRC letter to Nuclear Management Company LLC's Kewaunee Nuclear Power Plant, uKewaunee Nuclear Power Plant - Review of the Leak-Before-Break Evaluation for the Residual Heat Removal, 50

Enclosure 1 PG&E Letter DCL-03-183 Accumulator Injection Line, and Safety Injection System (TAC No. MB1301)," dated September 5, 2002.

24. NRC letter to Rochester Gas and Electric Company's R. E. Ginna Nuclear Power Plant, "Staff Review of the Submittal by Rochester Gas and Electric Company to Apply Leak-Before-Break Status to Portions of the R.E. Ginna Nuclear Power Plant Residual Heat Removal System Piping (TAC No. MA0389)," dated February 25,1999.

25. NRC letter to Northeast Nuclear Energy Company's Millstone Nuclear Power Station Unit 2, "Staff Review of the Submittal by Northeast Nuclear Energy Company to Apply Leak Before Break Status to the Pressurizer Surge Line, Millstone Nuclear Power Station Unit 2 (TAC No. MA4146)," dated May 4, 1999.

26. Westinghouse Electric LLC WCAP-14707, "Model 51 Steam Generator Limited Tube Support Plate Displacement Analysis for Dented or Packed Tube-to-Tube Support Plate Crevices," dated August 1996 (proprietary) and WCAP-14708, "Model 51 Steam Generator Limited Tube Support Plate Displacement Analysis for Dented or Packed Tube-to-Tube Support Plate Crevices," dated August 1996 (nonproprietary).

27. PG&E letter DCL-96-206, "Transmittal of WCAP-14707/14708,

'Model 51 Steam Generator Limited Tube Support Plate Displacement Analysis for Dented or Packed Tube to Tube Support Plate Crevices,"' dated October 4, 1996.

28. NRC letter to PG&E's Diablo Canyon Nuclear Power Plant, "Staff Review of WCAP-14707/14708, 'Model 51 Steam Generator Limited Tube Support Plate Displacement Analysis for Dented or Packed Tube-to-Tube Support Plate Crevices'- Diablo Canyon Power Plant, Units 1 and 2 (TAC Nos. M99011 and M99012)," dated January 10, 2000.

29. Draft NUREG/CR-xxxx, ANL-02/xx report "Sensitivity Studies of Failure of Steam Generator Tubes during Main Steam Line Break and Other Secondary Side Depressurization Events," dated December 2002, by S. Majumdar.

30. NUREG-1512, "Final Safety Evaluation Report Related to the Certification of the AP600 Standard Design, Docket No.52-003,"

dated August 1998.

31. NUREG-1462, Volume 1, "Final Safety Evaluation Report Related to the Certification of the System 80+ Design, Docket No.52-002,"

dated August 1994.

51

Enclosure 1 PG&E Letter DCL-03-183 7.2 Precedent Increased Voltage ODSCC ARC The NRC has previously approved the use of a 3-volt repair limit, which is higher than the repair limits under GL 95-05, for SG bobbin coil probe axial ODSCC indications constrained inside the thickness of a TSP for the South Texas Project Unit 2 plant in License Amendment No. 114 to Facility Operating License No. NPF-80, "South Texas Project (STP)

Unit 2 - Issuance of Amendment Revising the Technical Specifications to Implement 3-volt Alternate Repair Criteria for Steam Generator Tube Repair (TAC No. MA8271)," dated March 8, 2001. The NRC also has previously approved a similar 3-volt repair limit amendment for Cycle 6 of the Braidwood-1 in License Amendment No. 69 to Facility Operating License No. NPF-72 and for Cycle 8 of the Byron Nuclear Power Station Unit 1 in License Amendment No. 77 to Facility Operating License No.

NPF-37, "Issuance of Amendments (TAC Nos. M96498, M96499, M96500 and M96501)," dated November 9,1995 and for Cycle 7 of the Braidwood-1 in License Amendment No. 82 to Facility Operating License No. NPF-72, "Issuance of Amendments (TAC Nos. M91671, M91672, M91673 and M91674)," dated May 14,1997.

Except for the magnitude of the bobbin coil voltage repair limit and number of cycles for which the repair limit is applicable, the proposed 4-volt repair limit is similar to the 3-volt repair limits approved for South Texas Project Unit 2, Byron Unit 1, and Braidwood Unit 1. The magnitude of the proposed repair limit is 4-volts versus 3-volts which was approved for South Texas Project Unit 2, Byron Unit 1, and Braidwood Unit 1.

However, the 2 volt increase over the currently licensed GL 95-05 2-volt repair limit for the proposed 4-volt repair limit is the same as the 2 volt increase to the GL 95-05 1-volt repair limit which was approved for South Texas Project Unit 2, Byron Unit 1, and Braidwood Unit 1. The proposed 4-volt repair limit applies to multiple operating cycles versus the 3-volt repair limits approved for South Texas Project Unit 2, Byron Unit 1, and Braidwood Unit 1 that were only applicable for one operating cycle.

LBB Analysis The approach used to perform the LBB analysis for the MSLs for DCPP Units 1 and 2 is consistent with that used for LBB analyses approved by staff for the R.E. Ginna Nuclear Power Plant, the Millstone Nuclear Power Station Unit 2, and the Kewaunee Nuclear Power Plant. The LBB analyses for these plants were approved by staff in letters for Docket No.

50-244, "Staff Review of the Submittal by Rochester Gas & Electric to Apply Leak Before Break Status to Portions of R. E. Ginna Nuclear Power 52

Enclosure 1 PG&E Letter DCL-03-183 Plant Residual Heat Removal System Piping (TAC No. MA0389)," dated February 25,1999, Docket No. 50-336, "Staff Review of the Submittal by Northeast Nuclear Energy Company to Apply Leak Before Break Status to the Pressurizer Surge Line, Millstone Nuclear Power Station Unit 2 (TAC No. MA4146)," dated May 4, 1999, and Docket No. 50-305, "Kewaunee Nuclear Power Plant - Review of the Leak Before Break Evaluation for the Residual Heat Removal, Accumulator Injection Line, and Safety Injection System (TAC No. MB1301)," dated September 5, 2002, respectively.

The NRC has previously approved the application of LBB to main steam line piping inside containment for the AP600 Standard Design in NUREG-1512, uFinal Safety Evaluation Report Related to the Certification of the AP600 Standard Design, Docket No.52-003," dated August 1998, and for the System 80+ Design in NUREG-1462, Volume 1, "Final Safety Evaluation Report Related to the Certification of the System 80+ Design, Docket No.52-002," dated August 1994.

The NRC has previously approved application of LBB based on a 0.25 gpm leak rate detection capability for RCS leakage for Nuclear Management Company LLC's Kewaunee Nuclear Power Plant in the letter uKewaunee Nuclear Power Plant - Review of the Leak-Before-Break Evaluation for the Residual Heat Removal, Accumulator Injection Line, and Safety Injection System (TAC No. MB1301)," dated September 5, 2002, and for Rochester Gas and Electric Company's R. E. Ginna Nuclear Power Plant in the letter "Staff Review of the Submittal by Rochester Gas and Electric Company to Apply Leak-Before-Break Status to Portions of the R.E. Ginna Nuclear Power Plant Residual Heat Removal System Piping (TAC No. MA0389)," dated February 25,1999.

53

Enclosure 2 PG&E Letter DCL-03-183 Proposed Technical Specification Changes (mark-up)

Remove Page Insert Page 5.0-11 5.0-11 5.0-1 Ia 5.0-11a 5.0-11b 5.0-1 3a 5.0-13a 5.0-14 5.0-14 5.0-15 5.0-15 5.0-29 5.0-29 5.0-29a

There are no changes to this page. Page included for information only.Lnuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG) Tube Surveillance Program SG tube integrity shall be demonstrated by performance of the following augmented inservice inspection program.

The provisions of SR 3.0.2 are applicable to the SG Tube Surveillance Program test frequencies.

a. SG Sample Selection and Inspection - SG tube integrity shall be determined during shutdown by selecting and inspecting at least the minimum number of SGs specified in Table 5.5.9-1.

b. SG Tube Sample Selection and Inspection - The SG tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 5.5.9-2. The inservice inspection of SG tubes shall be performed at the frequencies specified in Specification 5.5.9.c and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 5.5.9.d. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all SGs; the tubes selected for these inspections shall be selected on a random basis except:

1. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;

2. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each SG shall include:

a) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),

b) Tubes in those areas where experience has indicated potential problems, c) A tube inspection (pursuant to Specification 5.5.9.d.1.h) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection, d) Indications left in service as a result of application of the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.

e) Tubes identified as W* tubes having a previously identified indication within the W* length shall be inspected using a rotating pancake coil (RPC) probe for the full length of the W* region during all future refueling outages. * **

(continued)

• Applicable for Units 1 and 2, Cycles 10, 11, 12, and 13 only

• • In-Situ Testing will be performed in accordance with PG&E letters DCL 98-148 dated October 22, 1998, and DCL 01-052 dated May 4, 2001, for Cycles 10 and 11 and letter DCL 01-095 dated September 13, 2001, for Cycles 12 and 13.

DIABLO CANYON - UNITS 1 & 2 5.0-10 Unit 1 - Amendment No. 435, 151 TAB 5.0 - R7 Unit 2 - Amendment No. 435, 151

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

3. The tubes selected as the second and third samples (if required by Table 5.5.9-2) during each inservice inspection may be subjected to a partial tube inspection provided:

a) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and b) The inspections include those portions of the tubes where imperfections were previously found.

4. Implementation of the steam generator tube/tube support plate repair criteria requires a 100% bobbin coil inspection for hot-leg and cold-leg support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The Insert I determination of the lowest cold-leg tube support plate intersection having Inser1 ODSCC indications shall be based on the performance of at least a 20%

andom sampling of tubes inspected over their full length.

5. Inspection of dented tube support plate intersections will be performed in accordance with WCAP-15573, Revision 1,to implement axial primary water stress corrosion cracking (PWSCC) depth-based repair criteria. The extent of required inspection is:

a) 100 percent bobbin coil inspection of all tube support plate (TSP) intersections.

b) Plus Point coil inspection of all bobbin coil indications at dented TSP intersections.

c) Plus Point coil inspection of all prior PWSCC indications left in service.

d) If bobbin coil is relied upon for detection of axial PWSCC in less than or equal to 2 volt dents, then on a SG basis perform Plus Point coil inspection of all TSP intersections having greater than 2 volt dents up to the highest TSP for which PWSCC has been detected in the prior two inspections or current inspection and 20% of greater than 2 volt dents at the next higher TSP. If a circumferential indication is detected in a dent of ax" volts in the prior two inspections or current inspection, Plus Point inspections will be conducted on 100% of dents greater than "x - 0.3" volts up to the affected TSP elevation in the affected SG, plus 20% of dents greater than "x - 0.3" volts at the next higher TSP. "x" is defined as the lowest dent voltage where a circumferential crack was detected.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-11 Unit 1 - Amendment No. 435, 452, TAB 5.0 - R7 Unit 2 - Amendment No. 435, 452,

There are no changes to this page. Page included for information only. l

.. als 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued) e) If bobbin coil is not relied upon for detection of axial PWSCC in less than or equal to 2 volt dents, then on a SG basis perform Plus Point coil inspection of all dented TSP intersections (no lower dent voltage threshold) up to the highest TSP for which PWSCC has been detected in the prior two inspections or current inspection and 20% of all dents at the next higher TSP.

f) For any 20% dent sample, a minimum of 50 dents at the TSP elevation shall be inspected. If the population of dents is less than 50 at the TSP elevation, then 100% of the dents at the TSP elevation shall be inspected.

The results of each sample inspection shall be classified into one of the following three categories:

Categorv Inspection Results Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.

c. Inspection Frequencies - The above required inservice inspections of SG tubes shall be performed at the following frequencies:

1. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; (continued)

DIABLO CANYON - UNITS 1 & 2 5.0-11a Unit 1 - Amendment No. 435,152 TAB 5.0 - R7 Unit 2 - Amendment No. 45,152

There are no changes to this page. Page included for information only.

5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

2. If the results of the inservice inspection of a SG conducted in accordance with Table 5.5.9-2 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 5.5.9.c.1. The interval may then be extended to a maximum of once per 40 months; and

3. Additional, unscheduled inservice inspections shall be performed on each SG in accordance with the first sample inspection specified in Table 5.5.9-2 during the shutdown subsequent to any of the following conditions:

a) Reactor-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.13; or b) A seismic occurrence greater than the Double Design Earthquake, or c) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or d) A main steam line or feedwater line break.

d. Acceptance Criteria

1. As used in this Specification:

a) Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications.

Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections; b) Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube; c) Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation; d)  % Degradation means the percentage of the tube wall thickness affected or removed by degradation.

e) Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective; f) Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness.

1) This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied. Refer to 5.5.9.d.1 .j for the repair limit applicable to these intersections.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-12 Unit 1 - Amendment No. 435 142 TAB 5.0 - R7 Unit 2 - Amendment No. 435 142

There are no changes to this page. Page included for information only.

s 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

2) This definition does not apply to the portion of the tube within the tubesheet below the W* length. Acceptable tube wall degradation within the W* length shall be defined as in 5.5.9.d.1.k. *

3) This definition does not apply to axial PWSCC indications, or portions thereof, which are located within the thickness of dented tube support plates which exhibit a maximum depth greater than or equal to 40 percent of the initial tube wall thickness. WCAP-1 5573, Revision 1, provides repair limits applicable to these intersections.

4) A tube which contains a tube support plate intersection with both an axial ODSCC indication and an axial PWSCC indication will be removed from service.

g) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of a Double Design Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 5.5.9.c.3, above; h) Tube Inspection means an inspection of the SG tube from the tube end (hot leg side) completely around the U-bend to the top support of the cold leg; i) Preservice Inspection means an inspection of the full length of each tube in each SG performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial Power Operation using the equipment and techniques expected to be used during subsequent inservice inspections; j) Tube Support Plate Plugging Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging limit is based on maintaining steam generator tube serviceability as described below:

(i) Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to the lower voltage repair limit (NOTE 1), will be allowed to remain in service.

(ii) Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (NOTE 1), will be repaired or plugged, except as noted in 5.5.9.d.lj (iii) below.

(continued)

• Applicable for Units I and 2,Cycles 10, 11, 12, and 13 only.

DIABLO CANYON - UNITS I & 2 5.0-13 Unit 1 - Amendment No. 435,4-14,152 TAB 5.0 - R7 Unit 2 - Amendment No. 435,454,152

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

(iii) Steam generator tubes, with indication of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (NOTE 1) but less than or equal to the upper voltage repair limit (NOTE 2), may remain in service if a rotating pancake coil inspection does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltane re air limit (NOTE

2) v~ irdThs specification i o plcbet Ftub plte suporntesecion atthefou loest hot-leg tube (continued)

DIABLO CANYON - UNITS 1 & 2 5.0-1 3a Unit 1 - Amendment No. 435,4514,452, TAB 5.0 - R7 Unit 2 - Amendment No. 435,454,452,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

(iv) Certain intersections as identified inWestinghouse letter to PG&E dated September 3, 1992, "Deformation of Steam Generator Tubes Following a Postulated LOCA and SSE Event", will be excluded from application of the voltage-based repair criteria as it is determined that these intersections may collapse or deform following a postulated LOCA +

SSE event.

(v) If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in 5.5.9.d.1lj (i),

5.5.9.d.1.j (ii), and 5.5.9.d.lj (iii)h-mid-cycle repair limits are determined from the following equations Insert 2 VMUL- VSL MURL (CL - At) 1.0 + NDE + Gr CL (CL-At VMLRL = VMURL - (VURL - VLRL) CL where:

VURL = upper voltage repair limit VLRL = lower voltage repair limit VMURL = mid-cycle upper voltage repair limit based on time into cycle VMLRL = mid-cycle lower voltage repair limit based on VMURL and time into cycle At = length of time since last scheduled inspection during which VURL and VLRL were implemented CL = cycle length (the time between two scheduled steam generator inspections)

VSL = structural limit voltage Gr = average growth rate per cycle length NDE = 95% cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20% has been approved by the NRC)

Implementation of these mid-cycle repair limits should follow the same approach as in TS 5.5.9.d.1lj (i), 5.5.9.d.1.j (ii), and 5.5.9.d.1lj (iii).

(continued)

DIABLO CANYON - UNITS I & 2 5.0-14 Unit 1 - Amendment No. 435, TAB 5.0 - R7 Unit 2 - Amendment No. 435,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

NOTE 1: The lower voltage repair limit is 2.0 volts for 7/8 inch diamet -

tubing at DCPP Units 1 and 2. lInsert 3 NOTE 2: The upper voltage repair limit is calculated according to the Insert 4 methodology in Generic Letter 95-05 as supplemented.

k) (*) W* Plugging Limit is used for disposition of an alloy 600 steam generator tube for continued service that is experiencing predominately axially oriented inside diameter stress corrosion cracking confined within the tubesheet, below the bottom of the WEXTEX transition (BWT). As used in this specification:

(i) Bottom of WEXTEX Transition (BWT) is the highest point of contact between the tube and tubesheet at, or below the top-of-tubesheet as determined by eddy current testing.

(ii) W* Length is the distance to the tubesheet below the BWT that precludes tube pull out in the event of the complete circumferential separation of the tube below the W* length. The W* length is conservatively set at: 1) an undegraded hot leg tube length of 5.2 inches for Zone A tubes and 7.0 inches for Zone B tubes, and 2) an undegraded cold leg tube length of 5.5 inches for Zone A tubes and 7.5 inches for Zone B tubes. Information provided in WCAP-14797, Revision 1, defines the boundaries of Zone A and Zone B.

(iii) Flexible W* Length is the W* length adjusted for any cracks found within the W* region. The Flexible W* Length is the total RPC-inspected length as measured downward from the BWT, and includes NDE uncertainties and crack lengths within W* as adjusted for growth.

(iv) W* Tube is a tube with equal to or greater than 40% degradation within or below the W* length that is left in service, and degraded within the limits specified in Specification 5.5.9d.1.k)(v).

(v) Within the tubesheet, the plugging (repair) limit is based on maintaining steam generator serviceability as described below:

1) For tubes to which the W* criteria are applied, the length of non-degraded tube below BWT shall be greater than or equal to the W*

length plus NDE uncertainties and crack growth for the operating cycle.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-15 Unit 1 - Amendment No. 435, TAB 5.0 - R7 Unit 2 - Amendment No. 435,

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.10 Steam Generator (SG) Tube Inspection Report

a. Within 15 days following the completion of each inservice inspection of SG tubes, the number of tubes plugged in each SG shall be reported to the Commission.

b. The complete results of the SG tube inservice inspection shall be submitted to the Commission in a report within 12 months following completion of the inspection. This Special Report shall include:

1) Number and extent of tubes inspected,

2) Location and percent of wall-thickness penetration for each indication of an imperfection, and

3) Identification of tubes plugged.

c. Results of SG tube inspections, which fall into Category C-3, shall be reported in a Special Report to the Commission within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

d. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC prior to returning the steam generators to service should any of the following arise:

1. If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution (reduced by estimated leakage by all other alternate repair criteria - *) exceeds the leak limit determined form the licensing basis dose calculation for the postulated main steamline break for the next operating cycle. Insert 5

2. If circumferential crack-like indications are detected at the tube support plate intersections.

3. If indications are identified that extend beyond the confines of the tube support plate.

4. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.

5. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10.2, notify the NRC and provide an assessment of the safety significance of the occurrence. sert 6 continued DIABLO CANYON - UNITS 1 & 2 5.0-29 Unit 1 - Amendment No. 4a5, TAB 5.0 - R7 Unit 2 - Amendment No. 435,

Enclosure 2 PG&E Letter DCL-03-183 Technical Specification Inserts Insert I a) For SGs that are locked by expansion joints in accordance with WCAP-16170-P, Revision 0, methodology (herein referred to as locked SGs), at tube support plate intersections other than at the four lowest hot-leg tube support plates, tubes with degradation attributed to axially oriented ODSCC within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (defined in 5.5.9.d.1.j) shall be inspected by Plus Point coil.

b) For locked SGs at tube support plate intersections at the four lowest hot-leg tube support plates, tubes with degradation attributed to axially oriented ODSCC within the bounds of the tube support plate with a bobbin voltage greater than 4 volts shall be inspected by Plus Point coil. An additional 100 tube intersections at the four lowest hot-leg tube support plates with degradation attributed to axially oriented ODSCC within the bounds of the tube support plate with a bobbin voltage less than 4 volts (100 total of all SGs, not necessarily selected at random) shall be inspected by Plus Point coil.

c) In each locked SG following the expansion process, 100% of the expanded tubes shall be Plus Point coil inspected at the top of tubesheet.

d) In each locked SG, following two complete cycles of operation and if the SGs are not to be replaced at the outage, 20% of the expanded tubes shall be Plus Point coil inspected at the top of tubesheet and expanded tube support plate locations. The Plus Point coil inspection shall be expanded to 100% of the expanded tubes in any SG that a circumferential indication greater than 100 degrees is found at the top of the tubesheet or above or below the tube support plate edges at tube support plate locations.

Insert 2 This specification is not applicable to tube support plate intersections at the four lowest hot-leg tube support plates in locked SGs.

Insert 3 For locked SGs, the lower voltage repair limit is 4.0 volts at tube support plate intersections at the four lowest hot-leg tube support plates and 2.0 volts at all other tube support plate intersections.

Enclosure 2 PG&E Letter DCL-03-183 Technical Specification Inserts (continued)

Insert 4 For locked SGs, the upper voltage repair limit is not applicable at tube support plate intersections at the four lowest hot-leg tube support plates.

Insert 5 For locked SGs, the estimated leakage from tube support plate intersections at the four lowest hot-leg tube support plates shall be calculated using the methodology of WCAP-16170-P, Revision 0, and combined with leakage from all other tube support plate intersections as calculated using the methodology of Generic Letter 95-05 modified to include a large leak rate assigned to indications predicted to burst in a Monte Carlo simulation as described in WCAP-1 6170-P, Revision 0.

Insert 6 For locked SGs, the burst probability is calculated using the methodology of WCAP-16170-P, Revision 0, that is, the axial tensile tearing rupture probability at tube support plate intersections at the four lowest hot-leg tube support plates is combined with the conditional burst probability at all other tube support plate intersections.

6. For locked SGs, if a significant increase in tube support plate ligament cracking is found in expanded tubes, an evaluation will be performed and the results reported to the NRC prior to restart.

7. For locked SGs, if circumferential indications are found in expanded tubes, an evaluation will be performed and the results reported to the NRC prior to restart.

Enclosure 3 PG&E Letter DCL-03-183 Proposed Technical Specification Changes (retyped)

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

3. The tubes selected as the second and third samples (if required by Table 5.5.9-2) during each inservice inspection may be subjected to a partial tube inspection provided:

a) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and b) The inspections include those portions of the tubes where imperfections were previously found.

4. Implementation of the steam generator tube/tube support plate repair criteria requires a 100% bobbin coil inspection for hot-leg and cold-leg support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersection having ODSCC indications shall be based on the performance of at least a 20%

random sampling of tubes inspected over their full length.

a) For SGs that are locked by expansion joints in accordance with WCAP-16170-P, Revision 0, methodology (herein referred to as locked SGs), at tube support plate intersections other than at the four lowest hot-leg tube support plates, tubes with degradation attributed to axially oriented ODSCC within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (defined in 5.5.9.d.1.j) shall be inspected by Plus Point coil.

b) For locked SGs at tube support plate intersections at the four lowest hot-leg tube support plates, tubes with degradation attributed to axially oriented ODSCC within the bounds of the tube support plate with a bobbin voltage greater than 4 volts shall be inspected by Plus Point coil. An additional 100 tube intersections at the four lowest hot-leg tube support plates with degradation attributed to axially oriented ODSCC within the bounds of the tube support plate with a bobbin voltage less than 4 volts (100 total of all SGs, not necessarily selected at random) shall be inspected by Plus Point coil.

c) In each locked SG following the expansion process, 100% of the expanded tubes shall be Plus Point coil inspected at the top of tubesheet.

d) In each locked SG, following two complete cycles of operation and if the SGs are not to be replaced at the outage, 20% of the expanded tubes shall be Plus Point coil inspected at the top of tubesheet and expanded tube support plate locations. The Plus Point coil inspection shall be expanded to 100% of the expanded tubes in any SG that a circumferential indication greater than 100 degrees is found at the top of the tubesheet or above or below the tube support plate edges at tube support plate locations.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-11 Unit 1 -Amendment No. 435, 452, TAB 5.0 - RX 11 Unit 2 - Amendment No. 435, 452,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

5. Inspection of dented tube support plate intersections will be performed in accordance with WCAP-1 5573, Revision 1, to implement axial primary water stress corrosion cracking (PWSCC) depth-based repair criteria. The extent of required inspection is:

a) 100 percent bobbin coil inspection of all tube support plate (TSP) intersections.

b) Plus Point coil inspection of all bobbin coil indications at dented TSP intersections.

c) Plus Point coil inspection of all prior PWSCC indications left in service.

d) If bobbin coil is relied upon for detection of axial PWSCC in less than or equal to 2 volt dents, then on a SG basis perform Plus Point coil inspection of all TSP intersections having greater than 2 volt dents up to the highest TSP for which PWSCC has been detected in the prior two inspections or current inspection and 20% of greater than 2 volt dents at the next higher TSP. If a circumferential indication is detected in a dent of "x" volts in the prior two inspections or current inspection, Plus Point inspections will be conducted on 100% of dents greater than "x - 0.3" volts up to the affected TSP elevation in the affected SG, plus 20% of dents greater than ux - 0.3" volts at the next higher TSP. ux" is defined as the lowest dent voltage where a circumferential crack was detected.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-1 1a Unit 1 - Amendment No. 435, 152 TAB5.0-RX 12 Unit 2 - Amendment No. 435,152

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued) e) If bobbin coil is not relied upon for detection of axial PWSCC in less than or equal to 2 volt dents, then on a SG basis perform Plus Point coil inspection of all dented TSP intersections (no lower dent voltage threshold) up to the highest TSP for which PWSCC has been detected in the prior two inspections or current inspection and 20% of all dents at the next higher TSP.

f) For any 20% dent sample, a minimum of 50 dents at the TSP elevation shall be inspected. If the population of dents is less than 50 at the TSP elevation, then 100% of the dents at the TSP elevation shall be inspected.

The results of each sample inspection shall be classified into one of the following three categories:

Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.

c. Inspection Frequencies - The above required inservice inspections of SG tubes shall be performed at the following frequencies:

1. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; (continued)

DIABLO CANYON - UNITS I & 2 5.0-11b Unit 1 -Amendment No. 435, 152 TAB 5.0 - RX 13 Unit 2 - Amendment No. 435, 152

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

(iii) Steam generator tubes, with indication of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (NOTE 1) but less than or equal to the upper voltage repair limit (NOTE 2), may remain in service if a rotating pancake coil inspection does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit (NOTE 2) will be plugged or repaired. This specification is not applicable to tube support plate intersections at the four lowest hot-leg tube support plates in locked SGs.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-13a Unit 1 - Amendment No. 1-5,454,452, TAB5.0-RX 16 Unit 2 - Amendment No. 435,454,4-2,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

(iv) Certain intersections as identified in Westinghouse letter to PG&E dated September 3, 1992, "Deformation of Steam Generator Tubes Following a Postulated LOCA and SSE Event", will be excluded from application of the voltage-based repair criteria as it is determined that these intersections may collapse or deform following a postulated LOCA +

SSE event.

(v) If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in 5.5.9.d.1j (i),

5.5.9.d.1.j (ii), and 5.5.9.d.1.j (iii). This specification is not applicable to tube support plate intersections at the four lowest hot-leg tube support plates in locked SGs. The mid-cycle repair limits are determined from the following equations:

VMURL SL (

1.0 + NDE+ Gr (CL-At)

CL (CL - At)

VMLRL = VMURL - (VURL - VLRL) CL where:

VURL = upper voltage repair limit VLRL = lower voltage repair limit VMURL = mid-cycle upper voltage repair limit based on time into cycle VMLRL = mid-cycle lower voltage repair limit based on VMURL and time into cycle At = length of time since last scheduled inspection during which VURL and VLRL were implemented CL = cycle length (the time between two scheduled steam generator inspections)

VSL = structural limit voltage Gr = average growth rate per cycle length NDE = 95% cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20% has been approved by the NRC)

Implementation of these mid-cycle repair limits should follow the same approach as in TS 5.5.9.d.lj (i), 5.5.9.d.1.j (ii), and 5.5.9.d.1.j (iii).

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-14 Unit 1 - Amendment No. 435, TAB 5.0 - RX 17 Unit 2 - Amendment No. 435,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

NOTE 1: The lower voltage repair limit is 2.0 volts for 7/8 inch diameter tubing at DCPP Units 1 and 2. For locked SGs, the lower voltage repair limit is 4.0 volts at tube support plate intersections at the four lowest hot-leg tube support plates and 2.0 volts at all other tube support plate intersections.

NOTE 2: The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented. For locked SGs, the upper voltage repair limit is not applicable at tube support plate intersections at the four lowest hot-leg tube support plates.

k) (*) W* Plugging Limit is used for disposition of an alloy 600 steam generator tube for continued service that is experiencing predominately axially oriented inside diameter stress corrosion cracking confined within the tubesheet, below the bottom of the WEXTEX transition (BWT). As used in this specification:

(i) Bottom of WEXTEX Transition (BWT) is the highest point of contact between the tube and tubesheet at, or below the top-of-tubesheet as determined by eddy current testing.

(ii) W* Length is the distance to the tubesheet below the BWT that precludes tube pull out in the event of the complete circumferential separation of the tube below the W* length. The W* length is conservatively set at: 1) an undegraded hot leg tube length of 5.2 inches for Zone A tubes and 7.0 inches for Zone B tubes, and 2) an undegraded cold leg tube length of 5.5 inches for Zone A tubes and 7.5 inches for Zone B tubes. Information provided in WCAP-14797, Revision 1, defines the boundaries of Zone A and Zone B.

(iii) Flexible W* Length is the W* length adjusted for any cracks found within the W* region. The Flexible W* Length is the total RPC-inspected length as measured downward from the BWT, and includes NDE uncertainties and crack lengths within W* as adjusted for growth.

(iv) W* Tube is a tube with equal to or greater than 40% degradation within or below the W* length that is left in service, and degraded within the limits specified in Specification 5.5.9d.l.k)(v).

(v) Within the tubesheet, the plugging (repair) limit is based on maintaining steam generator serviceability as described below:

1) For tubes to which the W* criteria are applied, the length of non-degraded tube below BWT shall be greater than or equal to the W*

length plus NDE uncertainties and crack growth for the operating cycle.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-15 Unit 1 - Amendment No. 435, TAB 5.0 - RX 18 Unit 2 - Amendment No. 435,

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.10 Steam Generator (SG) Tube Inspection Report

a. Within 15 days following the completion of each inservice inspection of SG tubes, the number of tubes plugged in each SG shall be reported to the Commission.

b. The complete results of the SG tube inservice inspection shall be submitted to the Commission in a report within 12 months following completion of the inspection. This Special Report shall include:

1) Number and extent of tubes inspected,

2) Location and percent of wall-thickness penetration for each indication of an imperfection, and

3) Identification of tubes plugged.

c. Results of SG tube inspections, which fall into Category C-3, shall be reported in a Special Report to the Commission within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

d. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC prior to returning the steam generators to service should any of the following arise:

1. If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution (reduced by estimated leakage by all other alternate repair criteria - *) exceeds the leak limit determined form the licensing basis dose calculation for the postulated main steamline break for the next operating cycle. For locked SGs, the estimated leakage from tube support plate intersections at the four lowest hot-leg tube support plates shall be calculated using the methodology of WCAP-16170-P, Revision 0, and combined with leakage from all other tube support plate intersections as calculated using the methodology of Generic Letter 95-05 modified to include a large leak rate assigned to indications predicted to burst in a Monte Carlo simulation as.described in WCAP-16170-P, Revision 0.

2. If circumferential crack-like indications are detected at the tube support plate intersections.

3. If indications are identified that extend beyond the confines of the tube support plate.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-29 Unit 1 -Amendment No. 435, TAB 5.0 - RX 33 Unit 2 - Amendment No. 435,

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.10 Steam Generator (SG) Tube Inspection Report

4. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.

5. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds I x 10-2, notify the NRC and provide an assessment of the safety significance of the occurrence. For locked SGs, the burst probability is calculated using the methodology of WCAP-16170-P, Revision 0, that is, the axial tensile tearing rupture probability at tube support plate intersections at the four lowest hot-leg tube support plates is combined with the conditional burst probability at all other tube support plate intersections.

6. For locked SGs, if a significant increase in tube support plate ligament cracking is found in expanded tubes, an evaluation will be performed and the results reported to the NRC prior to restart.

7. For locked SGs, if circumferential indications are found in expanded tubes, an evaluation will be performed and the results reported to the NRC prior to restart.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-29a Unit 1 - Amendment No. 425, TAB 5.0 - RX 34 Unit 2 - Amendment No. 435,

Enclosure 4 PG&E Letter DCL 03-183 Westinghouse Electric LLC WCAP 16170-NP, Revision 0, "Diablo Canyon SG Alternate Repair Criteria Based on Limited Tube Support Plate Displacement," dated November 2003 (nonproprietary)