RS-08-016, Application for Steam Generator Tube Interim Alternate Repair Criteria Technical Specification Amendment

Application for Steam Generator Tube Interim Alternate Repair Criteria Technical Specification Amendment

ML080560512
Person / Time
Site:	Braidwood
Issue date:	02/25/2008
From:	Simpson P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAW-08-2382, RS-08-016 LTR-CDME-08-11 NP
Download: ML080560512 (83)

Text

www.exelc)ncorp .com Exel6n Nuclear Exelon Generation 4300 Winfield Road Warrenville, Il 60555 10 CFR 50.90 RS-08-016 February 25, 2008 U . S . Nuclear Regulatory Commission Attn : Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos . 50-456 and 50-457

Subject:

Application for Steam Generator Tube Interim Alternate Repair Criteria Technical Specification Amendment

Reference:

Letter from P. R. Simpson (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Application for Steam Generator Tube Alternate Repair Criteria Technical Specification Amendment,"

dated November 29, 2007 In accordance with 10 CFR 50 .90, Exelon Generation Company, LLC (EGC) is requesting an amendment to Appendix A Technical Specifications (TS), of Facility Operating License Nos . NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2 . This amendment application proposes a one cycle revision to Technical Specification (TS) 5 .5 .9, "Steam Generator (SG) Program," to incorporate an interim alternate repair criteria in the provisions for SG tube repair criteria during Braidwood Station Unit 2 refueling outage 13 and the subsequent operating cycle. In addition, reporting requirement changes are proposed to TS 5.6.9, "Steam Generator (SG) Tube Inspection Report ." These changes are supported by Westinghouse Electric Company LLC, LTR-CDME-08-11 P-Attachment, "Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone ."

In a January 3, 2008, discussion between representatives of the Nuclear Energy Institute (NEI), U . S. Nuclear Regulatory Commission (NRC), and select Pressurized Water Reactor licensees, the NRC indicated that they would not be able to approve the referenced permanent revision to TS 5 .5.9 in time to support the Spring 2008 Braidwood Station Unit 2 refueling outage. In response to the NRC feedback, EGC is submitting this one-cycle amendment request.

February 25, 2008 U . S. Nuclear Regulatory Commission Page 2 This interim alternate repair criteria (IARC) amendment application does not supersede the referenced permanent amendment request. EGC requests that the NRC continue its review of the permanent ARC request subsequent to the review of this amendment application.

Although the proposed changes only affect Braidwood Station Unit 2, this submittal is being docketed for Braidwood Station Units 1 and 2 since the TS are common to Units 1 and 2 for Braidwood Station. contains information proprietary to Westinghouse Electric Company, LLC (Westinghouse) and is supported by an affidavit (Attachment 3) signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2 .390, "Public inspections, exemptions, requests for withholding ." Accordingly, it is requested that the information that is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR 2.390 . provides the non-proprietary version of the Attachment 4 information. provides a Westinghouse errata letter clarifying editorial changes to the and Attachment 5 documents . Attachment 7 provides EGC regulatory commitments related to tube end weld inspections at Braidwood Station Unit 2.

The attached request is subdivided as shown below.

Attachment 1 provides an evaluation of the proposed changes .

Attachment 2 includes the marked-up TS pages with the proposed changes indicated for Braidwood Station .

Attachment 3 provides an affidavit for withholding the proprietary information provided in Attachment 4. Also provided is the Westinghouse authorization letter, CAW-08-2382, "Application for Withholding Proprietary Information from Public Disclosure ."

Attachment 4 provides Westinghouse LTR-CDME-08-11 P-Attachment, "Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone," (Proprietary) .

Attachment 5 provides Westinghouse LTR-CDME-08-11 NP-Attachment, "Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone," (Nonproprietary) .

Attachment 6 provides Westinghouse LTR-CDME-08-25, "Errata for LTR-CDME 11 ; "Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone," dated February 12, 2008.

Attachment 7 provides EGC Regulatory Commitments related to tube end weld inspections to be performed with the implementation of the SG IARC TS amendment.

February 25, 2008 U . S . Nuclear Regulatory Commission Page 3 The proposed amendment has been reviewed by the Braidwood Station Plant Operations Review Committees and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program.

In accordance with 10 CFR 50.91(b), "State consultation," EGC is providing the State of Illinois with a copy of this letter and its non-proprietary attachments to the designated State Official .

EGC requests that this proposed license amendment change be approved by April 21, 2008, to support the inspection activities for Braidwood Station Unit 2, refueling outage 13. The requested implementation would be prior to the return to service from the Braidwood Station Unit 2 Spring 2008 refueling outage. If you have any questions about this letter, please contact Mr. David Chrzanowski at (630) 657-2816 .

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 25th day of February 2008.

R Patrick R . Simpson Manager - Licensing

bcc: Illinois Emergency Management Agency - Division of Nuclear Safety - w/o Attachments 3 and 4 Site Vice President - Braidwood Station - w/o Attachments 3, 4, 5, and 6 Regulatory Assurance Manager - Braidwood Station - w/o Attachments 3, 4, 5, and 6 Manager, Licensing - Braidwood, Byron and LaSalle Stations - w/o Attachments 3, 4, 5 and 6 Exelon Document Control Desk Licensing (Hard Copy)

Exelon Document Control Desk Licensing (Electronic Copy)

Commitment Tracking Coordinator - Cantera - w/o Attachments 3, 4, 5 and 6 Commitment Tracking Coordinator - Braidwood Station - w/o Attachments 3, 4, 5 and 6

Attachment 1 Evaluation of Proposed Changes Application for Steam Generator Tube Interim Alternate Repair Criteria Technical Specification Amendment Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. 50-456 and 50-457

Attachment 1 Evaluation of Proposed Changes INDEX 1 .0 DESCRIPTION

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

6.0 ENVIRONMENTAL CONSIDERATION

7 .0 PRECEDENT 8 .0 REFERENCES

Attachment 1 Evaluation of Proposed Changes 1 .0 DESCRIPTION This amendment application proposes a one cycle revision to Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program," for Braidwood Station to incorporate an interim alternate repair criteria (IARC) in the provisions for SG tube repair criteria during Braidwood Station Unit 2 (Braidwood 2) refueling outage 13 and the subsequent operating cycle. This IARC proposal requires full-length inspection of the SG tubes within the tubesheet but does not require plugging tubes if circumferential cracking observed in the region greater than 17 inches from the top of the tubesheet (TTS) is less than a value sufficient to permit the remaining circumferential ligament to transmit the limiting axial loads. This amendment application is required to preclude unnecessary plugging of SG tubes that maintain the necessary structural and leakage integrity. The proposed amendment also adds three additional reporting requirements to Braidwood Station TS 5.6.9, "Steam Generator (SG) Tube Inspection Report ."

Approval of this amendment application is requested to support Braidwood 2 inspection activities during refueling outage 13 in Spring 2008 and the subsequent operating cycle .

The approval of this interim ARC is requested as the existing one-cycle amendment expires at the end of the current operating cycle for Braidwood 2 (Reference 2).

2.0 PROPOSED CHANGE

EGC proposes to revise TS 5.5.9, "Steam Generator (SG) Program," by updating the current one-cycle ARC contained in the current wording in the Braidwood Station TS 5.6.9, "Steam Generator (SG) Tube Inspection Report," with a new one-cycle IARC .

Reference 2 approved one-cycle revisions to TS 5.5.9, for Braidwood Station Unit 2, to exclude from inspection and repair, that portion of the tube below 17 inches from the top of the tubesheet in the steam generators . Reference 3 proposed a permanent alternate repair criterion (ARC) to TS 5.5.9 to limit the inspection depth in the Braidwood 2 SG tube expansion zone, known as H*/B*. The H*/B* ARC seeks to minimize the depth of rotating coil inspection of the SG tubes within the tubesheet . The premise of H*/B* is that the expansion joint provides sufficient structural restraint to prevent the tube from pulling out of the tubesheet under normal operating and accident conditions, and that the accident induced leakage during accident conditions is bounded by a factor of two on the observed normal operating leakage.

In a January 3, 2008, discussion between representatives of the Nuclear Energy Institute (NEI), U . S. Nuclear Regulatory Commission (NRC), and select Pressurized Water Reactor licensees, the NRC indicated that they would not be able to approve the Reference 3 permanent revision to TS 5.5 .9 in time to support the Spring 2008 Braidwood Station Unit 2 refueling outage . In response to the NRC feedback, EGC is submitting this one-cycle amendment request. provides the technical justification for the IARC that requires full-length inspection of the tubes within the tubesheet, but does not require plugging tubes if the extent of any circumferential cracking observed in the region greater than 17 inches from the top of the tubesheet (TTS) is less than a value sufficient to permit the remaining circumferential ligament to transmit the limiting axial loads (i .e., the greater of 3 times the normal operating (NOP) or 1 .4 times the steam line break (SLB) end cap loads) .

Attachment 1 Evaluation of Proposed Changes Axial cracks below 17 inches from the TTS are not relevant to the tube pullout arguments because axial cracks do not degrade the axial load carrying capability of the tube . Axial cracks do not require plugging if they are below 17 inches from the TTS .

The changes from current TS wording are identified below with the affected sections italicized .

Braidwood Station TS 5 .5.9 c . 1 .

currently states : Braidwood Station proposed wording:

Tubes found by inservice inspection to Tubes found by inservice inspection to contain flaws in a non-sleeved region contain flaws in a non-sleeved region with a depth equal to or exceeding 40% with a depth equal to or exceeding 40%

of the nominal wall thickness shall be of the nominal wall thickness shall be plugged or repaired except if permitted plugged or repaired except if permitted to remain in service through application to remain in service through application of the alternate repair criteria discussed of the alternate repair criteria discussed in TS 5 .5 .9.c.4. For Unit 2 only, during in TS 5 .5.9.c.4 . For Unit 2 only, during Refueling Outage 12 and the Refueling Outage 13 and the subsequent operating cycle, flaws subsequent operating cycle, flaws identified in the portion of the tube from identified in the portion of the tube from the top of the hot leg tubesheet to the top of the tubesheet to 17 inches 17 inches below the top of the below the top of the tubesheet shall be tubesheet shall be plugged or repaired plugged or repaired upon detection .

upon detection .

Braidwood Station TS 5 .5 .9 c. 4. i . Braidwood Station proposed wording:

currently states :

For Unit 2 only, during Refueling For Unit 2 only, during Refueling Outage 12 and the subsequent Outage 13 and the subsequent operating cycle, flaws found in the operating cycle, flaws less than or equal portion of the tube below 17 inches from to 214 degree circumferential length the top of the hot leg tubesheet do not found in the portion of the tube below require plugging or repair. 17 inches from the top of the tubesheet do not require plugging or repair.

Tubes with axial indications found in the portion of the tube below 17 inches from the top of the tubesheet do not require plugging or repair.

Attachment 1 Evaluation of Proposed Changes Braidwood Station TS 5 .5 .9 d. currently Braidwood Station proposed wording:

states, in part :

Periodic SG tube inspections shall be Periodic SG tube inspections shall be performed. The number and portions of performed. The number and portions of the tubes inspected and methods of the tubes inspected and methods of inspection shall be performed with the inspection shall be performed with the objective of detecting flaws of any type objective of detecting flaws of any type (e.g., volumetric flaws, axial and (e.g ., volumetric flaws, axial and circumferential cracks) that may be circumferential cracks) that may be present along the length of the tube, present along the length of the tube, from the tube-to-tubesheet weld at the from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy at the tube outlet, and that may satisfy the applicable tube repair criteria . For the applicable tube repair criteria . The Unit 2 only, during Refueling Outage 12 tube-to-tubesheet weld is not part of the and the subsequent operating cycle, the tube.

portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded. The tube-to-tubesheet weld is not part of the tube .

Attachment 1 Evaluation of Proposed Changes Braidwood Station TS 5.6.9 currently Braidwood Station proposed j, k, and I does not have reporting requirements j, additions :

k, and I

j. For Unit 2, following completion of an inspection performed in Refueling Outage 13 (and any inspections performed in the subsequent operating cycle), the number of indications and location, size, orientation, and whether initiated on primary or secondary side for each flaw detected within the thickness of the tubesheet,

k. For Unit 2, following completion of an inspection performed in Refueling Outage 13 (and any inspections performed in the subsequent operating cycle), the operational primary to secondary leakage rate observed (greater than three gallons per day) in each steam generator (if it is not practical to assign the leakage to an individual steam generator, the entire primary to secondary leakage should be conservatively assumed to be from one steam generator) during the cycle preceding the inspection which is the subject of the report, and

l. For Unit 2, following completion of an inspection performed in Refueling Outage 13 (and any inspections performed in the subsequent operating cycle), the calculated accident leakage rate from the lowermost 4-inches of tubing for the most limiting accident in the most limiting steam generator.

Attachment 1 Evaluation of Proposed Changes As discussed above, EGC proposes to add three reporting requirements (i.e., TS 5.6.9 paragraphs j, k and I) to Braidwood Station TS 5.6.9 . Also editorial changes are proposed for Braidwood Station TS 5.5.9, paragraph c2i, adding "For Unit 2 only," and paragraph c.4, to change "may" to "shall ." These changes provide consistency with the wording throughout TS 5.5.9.

The marked-up TS pages provided in Attachment 2 indicate the specific wording changes for Braidwood Station.

3.0 BACKGROUND

Braidwood Station, Unit 2, contains four Westinghouse Model D5 recirculating, pre-heater type SGs . Each SG contains 4,570 thermally treated Alloy-600 U-tubes that have an outer diameter of 0.750 inch with a 0 .043-inch nominal wall thickness . The support plates are 1 .12 inches thick stainless steel and have quatrefoil broached holes. The tubing within the tubesheet is hydraulically expanded throughout the full thickness of the tubesheet. The tubesheet is approximately 21 inches thick . The low row U-bend region, up through row nine, received additional thermal stress relief following tube bending .

The unit operates on approximately 18-month fuel cycles .

The SG inspection scope is governed by: Braidwood Station TS 5 .5 .9; the Electric Power Research Institute (EPRI) Pressurized Water Reactor (PWR) SG Examination Guidelines ; regulatory documents and commitments; EGC ER-AP-420 procedure series (Steam Generator Management Program Activities) ; and the results of Braidwood 2 degradation assessment .

The inspection techniques and equipment are capable of reliably detecting the known and potential specific degradation mechanisms applicable to the Braidwood 2 SGs . The inspection techniques, essential variables and equipment are qualified to Appendix H, "Performance Demonstration for Eddy Current Examination," of the EPRI PWR SG Examination Guidelines .

In order to preclude unnecessarily plugging tubes in the Braidwood 2 SGs, an analysis was performed to identify the maximum flaw size in the bottom 4 inches of the tube within the tubesheet necessary to maintain structural and leakage integrity for both normal operating and accident conditions . The analysis (provided in Attachment

4) provides justification for an IARC that requires full-length inspection of the tubes within the tubesheet but does not require plugging tubes with axial indications or tubes with a certain arc length of circumferential cracking below 17 inches from the top of the tubesheet (i.e., the lower 4 inches) . If flaws are found within the top of tubesheet to 17 inches below the top of tubesheet, the tube must be repaired or removed from service. The IARC methodology was developed for the tubesheet region of Model D5 SGs considering the most stringent loads associated with plant operation, including transients and postulated accident conditions .

Attachment 1 Evaluation of Proposed Changes 4 .0 TECHNICAL ANALYSIS An evaluation has been performed by Westinghouse (provided as Attachment 4) to assess the need for removing tubes from service due to the occurrence of circumferentially or axially oriented cracks in a tubesheet. The conclusions of the evaluation are primarily threefold:

1 . Axial cracks in tubes below a distance of 17 inches below the top of the tubesheet can remain in service in the Braidwood 2 SGs as they are not a concern relative to tube pullout and leakage capability .

2. Circumferentially oriented cracks in tubes with an azimuthal extent of less than or equal to 214 degrees can remain in service for one cycle of operation .

3. Circumferentially oriented cracks in the tube-to-tubesheet welds with an azimuthal extent of less than or equal to 294 degrees can remain in service for one cycle of operation .

A bounding analysis approach is utilized for both the minimum ligament calculation and leakage ratio calculation . "Bounding" means that the most challenging conditions from the plants with hydraulically expanded Alloy 600TT tubing is used . The analysis includes the 0.75 inch diameter tube which is the tube size for the Braidwood 2 D5 model SGs . The most limiting conditions for structural evaluation depend on tube geometry and applied normal operating loads ; thus the conditions from the plant that result in the highest stress in the tube are used to define the minimum required circumferential ligament. The limiting leak rate ratio depends on the leak rate values assumed in the safety analysis and allowable normal operating leakage that results in the longest length of undegraded tube.

Discussion of Performance Criteria The performance criteria of the Nuclear Energy Institute (NEI) (Reference 4) are the basis for these analyses . The performance criteria, also provided in the Braidwood Station TS, are :

The structural integrity performance criterion is:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in

Attachment 1 Evaluation of Proposed Changes combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1 .0 on axial secondary loads.

The structural performance criterion is based on ensuring that there is reasonable assurance that a steam generator tube will not burst during normal operation or postulated accident conditions.

The accident-induced leakage performance criterion as stated in Braidwood Station TS 5.5.9 2. is:

The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm for all SGs.

Primary-to-secondary leakage is a factor in the dose releases outside containment resulting from a limiting design basis accident. The potential primary-to-secondary leak rate during postulated design basis accidents shall not exceed the control room or offsite radiological dose consequences required by 10 CFR 50 .67, "Accident source term," or GDC-19, "Control room," guidelines .

The IARC for the tubesheet region are designed to meet these criteria . The structural criterion regarding tube burst is inherently satisfied because the constraint provided by the tubesheet to the tube prohibits burst .

Limiting Structural Ligament Discussion As defined in the Westinghouse analysis provided as Attachment 4, the bounding remaining structural ligament which meets the Reference 4 Performance Criterion described above and required for the tube to transmit the operational loads is 115 degrees arc length . This assumes that the residual ligament is 100% of the tube wall in depth . A small circumferential initiating crack is predicted to grow to a throughwall condition before it is predicted to reach a limiting residual ligament . A residual ligament in a part-throughwall condition is not a significant concern, because of the assumption that all circumferential cracks detected are 100% throughwall.

Consideration of Non-Destructive Evaluation (NDE) Uncertaintv The NDE uncertainty must be addressed to assure that the as-indicated circumferential arc of the reported crack is a reliable estimate of the actual crack. The Electric Power Research Institute (EPRI) technique sheet ETSS 20510 .1 (Reference 5) describes the qualified technique used to detect circumferential PWSCC in the expansion transitions and in the tubesheet expansion zone (TEZ). The qualification data is provided in the ETSS .

Attachment 1 Evaluation of Proposed Changes The fundamental assumption for the IARC is that all circumferential cracks detected are 100% throughwall . Thus, even a shallow crack of small length will be considered to be throughwall . Further, tube burst is not an issue for the IARC because of the constraint provided by the tubesheet; rather, it is axial separation of the tube that is the principal concern. Assuming that all circumferential cracks are throughwall reduces the inspection uncertainty to the length of the cracks only. Further, the accuracy of the length determination is an issue only when the indicated crack approaches the allowable crack length (the complement of the required residual ligament) and if the indicated crack length is a reasonable estimate of the structural condition of the tube .

Prior investigations have correlated the axial strength of the tube to the Percent Degraded Area (PDA) of the flaw (Reference 6) . PDA takes into account the profile of the existing crack, including non-throughwall portions and shallow tails of the crack .

Using the data from Reference 5 for cracks with a 90%, or greater, throughwall condition from both NDE and destructive examination, a comparison of the actual crack lengths and corresponding PDA for the cracks to a theoretical PDA which assumes that cracks are 100% throughwall has been made . All of the points with a PDA of 60%, or greater fall below the theoretical PDA line . As the crack lengths increase, the separation of the actual PDA from the theoretical PDA tends to increase.

The conclusion that the as-indicated crack angle is conservative is further supported by considering the characteristics of the eddy current (EC) probes. As the probe traverses its path, a flaw will be detected as the leading edge of the field of view first crosses the location of the flaw, continuing until the trailing edge of the field of view passes the opposite end of the flaw. This is known as "lead-in" and "lead-out" of the probe and the effect of these are to render the indicated flaw length greater than the actual flaw length .

Therefore, it is concluded that the indicated flaw length will be conservative relative to the actual flaw length, especially when it is assumed that the entire length of the indicated flaw is 100% throughwall.

Based on the above, it is concluded that if the detected circumferential cracks are assumed to be 100% throughwall, the as-indicated crack lengths will be inherently conservative with respect to the structural adequacy of the remaining ligament.

Therefore, no additional uncertainty factor is necessary to be applied to the as-measured circumferential extent of the cracks .

Consideration of Crack Growth The growth of cracks due to PWSCC in this submittal request is dictated by four default growth rates from the Attachment 4 analysis . The distribution of growth rates is assumed to be log-normal . Typical values and conservative values are given, although it is recommended in Reference 7 to use the default values only when the historical information is not available and not to use the typical values unless the degradation is mild (no significant crack growth data exists for the circumferential cracking in the tubesheet expansion region .). Both sets provided in Attachment 4 have mean values and 95% upper bound values . For this analysis, the typical 95% upper bound growth rate is used .

Attachment 1 Evaluation of Proposed Changes The circumferential growth rates are expressed as inches per effective full power year (EFPY) .

Table 1 .0 Calculation of Required Minimum Ligament for 18 and 36 Months Operating Periods Bounding EFPY Growth Growth Growth for Minimum Critical Structural (In ./EFPY) (Deg ./EFPY) Operating Structural Ligament Ligament (1) (2) Period Ligament (degrees)

(degrees arc (degrees arc length) length 18 Calendar Tube Month (CM) 1 .5 .12 20.65 31 115 146 Operation 36 CM Operation 3.0 .12 20 .65 62 115 177

1) 95% upper value of typical growth rates from the Attachment 4 analysis 2 Based on smallest mean tubesheet bore dimension for the limiting SG design i .e., Model F The residual structural ligament must be adjusted for growth during the anticipated operating period between the current and the next planned inspection . Typically, the operating periods for Braidwood 2 is 18 calendar months (1 .5 EFPY) . For the Braidwood 2 SGs, referring to Table 1 .0 above, the maximum allowable throughwall circumferential crack size in a SG tube is 214° (=360° -146 °) for one cycle of operation .

No additional uncertainty factor is necessary to be applied to the as-measured circumferential extent of the cracks .

Primary-to-Secondary Leakaqe Discussion A basis, using the D'Arcy formula for flow through a porous medium, is provided to assure that the accident induced leakage for the limiting accident will not exceed the value assumed in the safety analysis for the plant if the observed leakage during normal operation is within its limits for the bounding plant is discussed in Attachment 4. The bounding plant envelopes all plants who are candidates for applying H*/B*. The D'Arcy formulation was previously compared to other potential models such as the Bernoulli equation or orifice flow formulation and was found to provide the most conservative results .

Assuming zero contact pressure in the tube joint, the length of undegraded crevice required to limit the accident induced leakage to less than the value assumed in the safety analysis for the limiting plant is calculated to be 3.78 inches . By definition of the IARC, a tube that can remain in service has an undegraded crevice of 17 inches .

Therefore, a factor of safety of 4.5 is available (17 inches /3.78 inches). Expressed in length terms, the length margin in the crevice is 13.22 inches . Significant margin on crevice length is available even if only the distance below the neutral axis of the tubesheet is considered . This distance is approximately 6.5 inches . A factor of safety of 1 .72 is available. Expressed in length terms, the length margin in the crevice is 2.72 inches below the neutral axis of the tubesheet . During normal operating conditions, the tubesheet flexes due to differential pressure loads, causing the tubesheet holes above the neutral axis to dilate, and below the neutral axis, to constrict. No mechanical benefit is assumed in the analysis due to tubesheet bore constriction below the neutral axis of 10 of 16

Attachment 1 Evaluation of Proposed Changes the tubesheet; however, first principles dictate that the tubesheet bore and crevice must decrease . Therefore, the leakage analysis provided is conservative .

Based on the above, with a length of undegraded crevice of 17 inches, it is concluded that, for the limiting plant, if the normal operating leakage is within its allowable value, the accident-induced leakage will also be within the value assumed in the safety analysis .

Inspection and Repair of Tube The tube below the IARC depth (i .e., 17 inches below the TTS) will be examined with a qualified technique, e.g., +PointTM probe . Axial flaws have no impact on the structural integrity of the tube in this region and may be left in service. Circumferential indications which exceed the maximum acceptable tube flaw size of 214 degrees will be plugged .

Flaws that require plugging will result in expansion per EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines ." Stress concentration areas may be used to define the extent of the expansion, e.g ., if a repairable indication is located in a tube bulge (BLG) or overexpansion (OXP), the expansion may be limited to the non-inspected BLG/OXPs.

Inspection and Repair of Tube End Welds For the tube end weld eddy current examination, the weld will be evaluated on a best effort basis for crack-like indications . Crack-like indications in the tube end weld that exceed the maximum acceptable weld flaw size of 294 degrees will be visually examined on a best effort basis with inspection systems capable of achieving a resolution similar to the Maximum Procedure Demonstration Lower Case Character Height as discussed in ASME Section XI . Dose rates in steam generator channel heads will typically necessitate performance of the inspection using remote inspection systems.

If the visual examination confirms that a flaw is present in the weld-to-tube interface and exceeds the maximum acceptable weld flaw size of 294 degrees, the tube will be plugged. This approach is conservative based on the assumption that the full visible length of indications are throughwall.

Tubes containing visual indications in the tube end weld that exceed the acceptance criteria will be removed from service by tube plugging . The installed plug joint becomes the pressure boundary . Plug installation results in significant contact pressure between the tube and the plug, as well as the tube and the tubesheet. The high contact pressures in the expanded plug load path will limit leakage, such that it would be indistinguishable from indications left inservice in the length of the tube below 17 inches from the top of the tubesheet and is, therefore, bounded by the leakage analysis for the IARC .

If the visual examination confirms that the weld-to-tube joint is sound and a flaw exists in the weld-to-clad joint, the tube is also acceptable to leave in service. Leaving the tube in service is acceptable because the attached weld prevents the tube from being ejected in an accident scenario .

Attachment 1 Evaluation of Proposed Changes An evaluation for tube end welds that require plugging will be completed under the Corrective Action Program to provide reasonable assurance that unacceptable welds are removed from service.

5.0 REGULATORY ANALYSIS

Steam Generator (SG) tube inspection and repair limits are specified in Section 5.5.9, "Steam Generator (SG) Program," of the Braidwood Station Technical Specifications (TS) . The current TS require that flawed tubes be repaired if the depths of the flaws are greater than or equal to 40 percent throughwall. The TS 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 General Design Criteria (GDC) 14, 15, 30, 31, and 32 of 10 CFR 50, Appendix A.

Specifically, the GDC state that the Reactor Coolant Pressure Boundary (RCPB) shall have "an extremely low probability of abnormal leakage . . . and gross rupture" (GDC 14), "shall be designed with sufficient margin" (GDCs 15 and 31), shall be of "the highest quality standards practical" (GDC 30), and shall be designed to permit "periodic inspection and testing . . . to assess . . . structural and leaktight integrity" (GDC 32).

Structural integrity refers to maintaining adequate margins against gross failure, rupture, and collapse of the steam generator tubing . Leakage integrity refers to limiting primary to secondary leakage during all plant conditions to within acceptable limits .

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 NRC 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 with the implementation of the interim alternate repair criteria discussed above.

5.1 NO SIGNIFICANT HAZARDS CONSIDERATION The proposed amendment is for an interim alternate repair criterion (IARC) that requires full-length inspection of the tubes within the tubesheet but does not require plugging tubes if any circumferential flaws observed in the region greater than 17 inches from the top of the tubesheet (TTS), that is, the lower 4 inches of the tube within the tubesheet is less than a value sufficient to permit the remaining circumferential ligament to transmit the limiting axial loads . In addition, since axial flaws below 17 inches from the TTS do not degrade the axial load carrying capability of the tube, axial flaws do not require plugging if they are below 17 inches from the TTS.

The proposed amendment also adds three reporting requirements to Braidwood Station TS 5.6.9, "Steam Generator (SG) Tube Inspection Report ."

EGC 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:

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

Response: No.

Of the various accidents previously evaluated, the proposed changes only affect the steam generator tube rupture (SGTR), postulated steam line break (SLB), locked rotor and control rod ejection accident evaluations. Loss-of-coolant accident (LOCA) conditions cause a compressive axial load to act on the tube. Therefore, since the LOCA tends to force the tube into the tubesheet rather than pull it out, it is not a factor in this amendment request. Another faulted load consideration is a safe shutdown earthquake (SSE); however, the seismic analysis of Model D5 steam generators has shown that axial loading of the tubes is negligible during an SSE.

At normal operating pressures, leakage from primary water stress corrosion cracking (PWSCC) below 17 inches from the top of the tubesheet is limited by both the tube-to-tubesheet crevice and the limited crack opening permitted by the tubesheet constraint .

Consequently, negligible normal operating leakage is expected from cracks within the tubesheet region .

For the SGTR event, the required structural margins of the steam generator tubes is maintained by limiting the allowable ligament size for a circumferential crack to remain in service to 214 degrees below 17 inches from the top of the tubesheet. Tube rupture is precluded for cracks in the hydraulic expansion region due to the constraint provided by the tubesheet. The potential for tube pullout is mitigated by limiting the allowable crack size to 214 degrees, which takes into account eddy current uncertainty and crack growth rate . It has been shown that a circumferential crack with an azimuthal extent of 214 degrees meets the performance criteria of NEI 97-06, Rev. 2, "Steam Generator Program Guidelines" and the Draft Regulatory Guide (RG) 1 .121, "Bases for Plugging Degraded PWR Steam Generator Tubes ." Likewise, a visual inspection will be conducted as necessary to confirm that a circumferential crack of greater than 294 degrees does not remain in service in the tube-to-tubesheet weld metal in any tube thereby mitigating the potential for tube pullout. Therefore, the margin against tube burst/pullout is maintained during normal and postulated accident conditions and the proposed change does not result in a significant increase in the probability or consequence of a SGTR.

The probability of a SLB is unaffected by the potential failure of a SG tube as the failure of a tube is not an initiator for a SLB event. SLB leakage is limited by leakage flow restrictions resulting from the leakage path above potential cracks through the tube-to-tubesheet crevice. The leak rate during postulated accident conditions has been shown to remain within the accident analysis assumptions for all axial or circumferentially oriented cracks occurring 17 inches below the top of the tubesheet. Since normal operating leakage is limited to 0.10 gallons per minute (gpm) (or 150 gallons per day (gpd)), the attendant accident condition leak rate, assuming all leakage to be from indications below 17 inches from the top of the tubesheet would be bounded by 0.5 gpm .

This value is within the accident analysis assumptions for the limiting design basis accident for Braidwood 2, which is the postulated SLB event.

Attachment 1 Evaluation of Proposed Changes Based on the above, the performance criteria of NEI-97-06, Rev. 2 and RG 1 .121 continue to be met and 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 kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not introduce any changes or mechanisms that create the possibility of a new or different kind of accident. Tube bundle integrity is expected to be maintained for all plant conditions upon implementation of the interim alternate repair criteria . The proposed change does not introduce any new equipment or any change to existing equipment. No new effects on existing equipment are created nor are any new malfunctions introduced .

Therefore, based on the above evaluation, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response : No.

The proposed change maintains the required structural margins of the steam generator tubes for both normal and accident conditions . NEI 97-06, Rev. 2 and RG 1 .121 are used as the basis in the development of the interim alternate repair criteria (IARC) methodology for determining that steam generator tube integrity considerations are maintained within acceptable limits . RG 1 .121 describes a method acceptable to the NRC staff for meeting General Design Criteria 14, 15, 31, and 32 by reducing the probability and consequences of an SGTR . RG 1 .121 concludes that by determining the limiting safe conditions of tube wall degradation beyond which tubes with unacceptable cracking, as established by inservice inspection, should be removed from service or repaired, the probability and consequences of a SGTR are reduced . This RG uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the ASME Code.

For axially oriented cracking located within the tubesheet, tube burst is precluded due to the presence of the tubesheet. For circumferentially oriented cracking in a tube or the tube-to-tubeshet weld, the Westinghouse analysis, provided in report "LTR-CDME-08-11 P-Attachment," defines a length of remaining tube ligament that provides the necessary resistance to tube pullout due to the pressure induced forces (with applicable safety factors applied) . Additionally, it is shown that application of the IARC will not result in unacceptable primary-to-secondary leakage during all plant conditions, including transients and postulated accident conditions.

Based on the above, it is concluded that the proposed changes do not result in any reduction in a margin of safety .

Based on the above, EGC concludes that the proposed amendment 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.

14 of 16

Attachment 1 Evaluation of Proposed Changes 5.2 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TS as part of the license. The Commission's regulatory requirements related to the content of the TS are contained in Title 10, Code of Federal Regulations (10 CFR), Section 50 .36, "Technical specifications . The TS requirements in 10 CFR 50.36 include the following categories : (1) safety limits, limiting safety systems settings and control settings, (2) limiting conditions for operation (LCO), (3) surveillance requirements, (4) design features, and (5) administrative controls . The SG tube inspection requirements are included in the TS in accordance with 10 CFR 50.36(c)(5),

"Administrative Controls."

As stated in 10 CFR 50.59, "Changes, tests, and experiments," paragraph (c)(1)(i), a licensee is required to submit a license amendment pursuant to 10 CFR 50.90, "Application for amendment of license or construction permit," if a change to the TS is required . Furthermore, the requirements of 10 CFR 50 .59 necessitate that the NRC approve the TS changes before the TS changes are implemented. EGC's submittal revising the requirements of TS 5.5.9, "Steam Generator Program" to allow circumferentially oriented flaws in tubes with less than or equal to 214 degrees in extent and axial flaws in tubes below a distance of 17 inches below the top of the tubesheet to remain in service as well as the changes to the Braidwood Station TS 5.6.9, "Steam Generator (SG) Tube Inspection Report," meet the requirements of 10 CFR 50.59(c)(1)(i) and 10 CFR 50.90.

Draft RG 1 .121 margins against burst are maintained for both normal and postulated accident conditions due to the constraint provided by the tubesheet.

6 .0 ENVIRONMENTAL CONSIDERATION EGC has evaluated the proposed amendment for environmental considerations . The review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, and would change an inspection or surveillance requirement.

However, 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 amendments meet the eligibility criterion for categorical exclusion set for in 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 PRECEDENT None .

Attachment 1 Evaluation of Proposed Changes

8.0 REFERENCES

(1) Letter from D. M. Benyak (Exelon Generation Company, LLC) to U. S. NRC, "Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification," dated August 18, 2006 (2) Letter from R. F. Kuntz (U . S. NRC) to C. M. Crane (Exelon Generation Company, LLC), "Braidwood Station, Unit 2 - Issuance of Amendments Re : Steam Generator Inspection Criteria (TAC No. MC8969)," dated October 24, 2006 Letter from P. R . Simpson (Exelon Generation Company, LLC) to U . S. Nuclear Regulatory Commission, "Application for Steam Generator Tube Alternate Repair Criteria Technical Specification Amendment," dated November 29, 2007 (4) NEI 97-06, Rev. 2, "Steam Generator Program Guidelines," May 2005 (5) ETSS #20510 .1, "Technique for Detection of Circumferential PWSCC at Expansion Transitions" (6) EPRI TR-107197, "Depth Based Structural Analysis Methods for Steam Generator Circumferential Indications," November 1997 EPRI 1012987, "Steam Generator Integrity Assessment Guidelines," July 2006

Attachment 2 Application for Steam Generator Tube Alternate Repair Criteria Technical Specification Amendment Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos . 50-456 and 50-457 Braidwood Station Marked-up Technical Specifications Pages 5.5-8 5.5-9 5.6-7

Attachment 2 Application for Steam Generator Tube Alternate Repair Criteria Technical Specification Amendment Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. 50-456 and 50-457 Braidwood Station Marked-up Technical Specifications Pages 5.5-8 5 .5-9 5 .6-7

Programs and Manuals 5 .5 5 .5 Programs and Manuals 5 .5 .9 Steam Generator (SG) Program (continued)

2. Accident induced leakage performance criterion : The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG .

Leakage is not to exceed a total of 1 gpm for all SGs .

3. The operational LEAKAGE performance criteria is specified in LCO 3 .4 .13, "RCS Operational LEAKAGE ."

c. Provisions for SG tube repair criteria .

1. Tubes found by inservice inspection to contain flaws in a non-sleeved region with a depth equal to or exceeding 40% of the nominal wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in TS 5 .5 .9 .c .4 . For Unit 2 only, during Refueling Outage 13 and the subsequent operating cycle, flaws identified in the portion of the tube from the top of the tubesheet to 17 inches below the top of the tubesheet shall be plugged or repaired upon detection .

2. Sleeves found by inservice inspection to contain flaws with a depth equal to or exceeding the following percentages of the nominal sleeve wall thickness shall be plugged :

i . For Unit 2 onl ,r , TIG welded sleeves (per TS 5 .5 .9 .f .2 .i) : 32%

3. Tubes with a flaw in a sleeve to tube joint that occurs in the sleeve or in the original tube wall of the joint shall be plugged .

4. The following tube repair criteria shall be applied as an alternate to the 40% depth-based criteria of Technical Specification 5 .5 .9 .c .1 :

i . For Unit 2 only, during Refueling Outage 13 and the subsequent operating cycle, 00.

BRAIDWOOD - UNITS 1 & 2 5 .5 - 8 Amendment

Programs and Manuals 5 .5 5 .5 Programs and Manuals 5 .5 .9 Steam Generator (SG) Program (continued)

d. Provisions for SG tube inspections . Periodic SG tube inspections shall be performed . The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e .g ., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria .

g 4w~eshee4 is eor-l~QPO~ The tube-to-tubesheet weld is not part of the tube . In addition to meeting the requirements of d .l, d .2, and d .3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection . An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations .

l. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement .

2. Inspect 100% of the Unit 1 tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months . The first sequential period shall be considered to begin after the first inservice inspection of the SGs . In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period . No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected .

Inspect 100% of the Unit 2 tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months . The first sequential period shall be considered to begin after the first inservice inspection of the SGs . In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period . No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected .

BRAIDWOOD - UNITS 1 & 2 5 .5 - 9 Amendment

INSERT A flaws less than or equal to 214 degree circumferential length found in the portion of the tube below 17 inches from the top of the tubesheet do not require plugging or repair . Tubes with axial indications found in the portion of the tube below 17 inches from the top of the tubesheet do not require plugging or repair .

Reporting Requirements 5 .6 5 .6 Reporting Requirements 5 .6 .8 Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported in the Inservice Inspection Summary Report in accordance with 10 CFR 50 .55a and ASME Section XI, 1992 Edition with the 1992 Addenda .

5 .6 .9 Steam Generator (SG) Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5 .5 .9, Steam Generator (SG) Program .

The report shall include :

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged or repaired to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing,

h. The effective plugging percentage for all plugging and tube repairs in each SG, i . Repair method utilized and the number of tubes repaired by each repair method:,

Insert B BRAIDWOOD - UNITS 1 & 2 5 .6 - 7 Amendment

Irs 4r' t B j . For Unit 2, following completion of an inspection performed in Refueling Outage 13 (and any inspections performed in the subsequent operating cycle), the number of indications and location, size, orientation, and whether initiated on primary or secondary side for each flaw detected within the thickness of the tubesheet, k . For Unit 2, following completion of an inspection performed in Refueling Outage 13 (and any inspections performed in the subsequent operating cycle), the operational primary to secondary leakage rate observed (greater than three gallons per day) in each steam generator (if it is not practical to assign the leakage to an individual steam generator, the entire primary to secondary leakage should be conservatively assumed to be from one steam generator) during the cycle preceding the inspection which is the subject of the report, and l . For Unit 2, following completion of an inspection performed in Refueling Outage 13 (and any inspections performed in the subsequent operating cycle), the calculated accident leakage rate from the lowermost 4-inches of tubing for the most limiting accident in the most limiting steam generator .

Attachment 3 Application for Steam Generator Tube Alternate Repair Criteria Technical Specification Amendment Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Dock et Nos. 50-456 and 50-457 Westinghouse Authorization Letter CAW-08-2382 Application for Withholding Proprietary Information from Public Disclosure

Westinghouse Westinghouse Electric Company Nuclear Services P.O . Box 355 Pittsburgh, Pennsylvania 15230-0355 USA U .S . Nuclear Regulatory Commission Directtel : (412)'174-4643 Document Control Desk Directfax : (412)374-401l Washington, DC 20555-0001 e-mail : greshaja@,westingh o use. corn Our ref CAW-08-2382 February 6, 2008 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE Subject : LTR-CDME-08-11 P-Attachment, "Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Gone," dated January 31, 2008 (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-08-2382 signed by the owner of the proprietary information, Westinghouse Electric Company LLC . The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2 .390 of the Commission's regulations .

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Exelon Generation Company, LLC Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-08-2382, and should be addressed to J . A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355 .

Very truly yours,

!` J .A. Gresharn, Manager 61 Regulatory Compliance and Plant Licensing Enclosures cc : Jon Thompson (NRC O-7E I A)

CAW-08-2382 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF ALLEGHENY :

Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

i

. A. Gresham, Manager Regulatory Compliance and Plant Licensing Sworn to and subscribed before me this 6"' day of February, 2008 Notary Public COMMONW~ALTW CF pENNSYI.VANIA Notarial Seal Sharon L Marble, Notary Public Motx0eville Boro, Allegheny County Ny Commission Expines Jan. 29, 20 11 I Member, Pennsylvania Association of Notaries

2 CAW-08-2382 I am Manager, Regulatory Compliance and Plant Licensing, in Nuclear Services, Westinghouse Electric Company LL,C (Westinghouse), and as such, l have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) 1 am making this Affidavit in conformance with the provisions of 10 CFR Section 2 .390 of the Commission's regulations and in conjunction with the Westinghouse "Application for Withholding" accompanying this Affidavit .

1 have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information .

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2 .390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld .

The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public . Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence . The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows :

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc .) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies .

3 CAW-08-2382 (b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability .

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product .

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers .

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse .

It contains patentable ideas, for which patent protection may be desirable .

There are sound policy reasons behind the Westinghouse system which include the following :

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors . It is, therefore, withheld from disclosure to protect the Westinghouse competitive position .

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information .

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense .

(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage . If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage .

4 CAW-08-2382 (e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries .

The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage .

(iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in LTR-CDME-08-11 P-Attachment, "Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone,"

dated January 31, 2008 (Proprietary), for submittal to the Commission, being transmitted by Exelon Generation Company, LLC Application for Withholding Proprietary Information from Public Disclosure to the Document Control Desk. The proprietary information as submitted for use by Westinghouse for Byron Unit 2 and Braidwood Unit 2 is expected to be applicable to other licensee submittals in support of implementing an interim alternate repair criterion (IARC) that requires a full-length inspection of the tubes within the tubesheet but does not require plugging tubes with a certain arc length of circumferential cracking below 17 inches from the top of the tubesheet .

This information is part of that which will enable Westinghouse to :

(a) Provide documentation of the analyses, methods, and testing for the implementation of an interim alternate repair criterion for the portion if the tubes within the tubesheet of the Byron Unit 2 and Braidwood Unit 2 steam generators .

5 CAW-08-2382 (b) Assist the customer in obtaining NRC approval of the Technical Specification changes associated with the interim alternate repair criterion.

Further this information has substantial commercial value as follows :

(a) Westinghouse plans to sell the use of similar information to its customers for the purposes of meeting NRC requirements for licensing documentation .

(b) Westinghouse can sell support and defense of the technology to its customers in the licensing process .

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar calculation, evaluation and licensing defense services for commercial power reactors without commensurate expenses . Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information .

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

Attachment 5 Application for Steam Generator Tube Alternate Repair Criteria Technical Specification Amendment Braidwood Station, Units 1 and 2 Facility Operating License Nos . NPF-72 and NPF-77 NRC Docket Nos . 50-456 and 50-457 Westinghouse LTR-CDME-08-11 NP-Attachment Non-Proprietary Version

WESTINGHOUSE NON-PROPRIETARY CLASS 3 LTR-CDME-08-11 NP-Attachment Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone January 31, 2008 Westinghouse Electric Company LLC P.O . Box 158 Madison, PA 15663

© 2008 Westinghouse Electric Company LLC All Rights Reserved

TABLE OF CONTENTS 1 .0 Introduction .. . .. . .. ... ... . .. ...... . .. ... ... ... .. . ... ... .. . .. ... . .. . .. ... ... . .. ... ... ... ... . .. . .. ... . .. . .. ... ... .. . .. ... ...1 2.0 Performance Criteria ... ... . .. ... . .. . .. ... ... ... ... ... ... .. . .. ... ... ... ... ... ... ... ... .. . .. . ... . . . .. .... .. . ... .. . .. ... ... ... 3 2.1 ReferenceS .. .. . ... ... ... ... ... .. . ... ... .. . .. ... ... .. . .. . .. . .. . ...... .. . .. ... ... . .. ... ... . .. ... ... . .. . .. ... ... .. . .. . 4 3 .0 Structural Evaluation for Minimum Circumferential Ligament. . ... ... .. . .. .... .. . ..... ... ... .. . .. . 5 3.1 Introduction . . . . ... ... .. . ... ... .. . .. . . . .. . . .. ... . .. .. . .. . ... .. . .. ... . .. . .. ... ... . .. .. . . .. . . . . . .. .. . . . .. . .. . ... .. . .. . 5 3.2 Analysis. ... ... ... ... ... .. . .. . ... .. . .. ... ... . .. ... ... .. . .. ...... . .. ... ... . .. ... ... ... ... ... ... ... ... ... .. . .. ... ... .. . .5 3 .2 .1 Description of the Steam Generator Models .. ... ... ... ... ... ... ... ... .. . .. ... ... .. . . 5 3 .2 .2 Flaw Geometries . .. ... ... .. . .. ...... . .. ... ... ... .. . .. . ... .. . ... ... .. . .. . ... .. . .. ... ... . .. ... ... .. .. 5 3 .2 .3 Initiation .. .. . .. . ... .. . .. ... ... .. . .. ... ... ... ... ... ... .. . ... ... .. . .. ... . .. . ... .. ... . .. .... .. ... .. ... . .. . . 6 3 .2 .4 Pressure Loading for Flaws in the Tube Wall .. . ... .. . ... ..... . .. ... ... .. . .. ... ... .. 6 3 .2 .5 Pressure Loading Effects in the Weld Metal. . .. .... .. . .. ...... . .. ... ... .. . .. ... ... .. 7 3.2 .6 Constraint.. . .. ... ... .. . ... ... .. . .. ... ... ... .. . .. .... .. . .. ... ... .. . ... ... . .. ... ... . .. ... ... .. ... ... ... ..7 3.2 .7 Force Balance.. . .. ... ........ ... ... ... ... .. . .. ... . .. . .. ... ... .. . ... ... ... ..... . .. . .. ...... ... .. . ..... 8 3.3 Results and Discussion . ... .. . ..... ... ... ... ... ... .. . .. .... .. . .. ... ... ... ... ... .. . ...... .. . .. ... ... ... .. . .. . 10 3.3 .1 Steam Generator Tube Wall Cross-Section .. .. . .. .... .. . .. ... . .. . .. ... ... ... .... . . 11 3.3 .2 Steam Generator Tube Cross-Section with an Initial 40 Degree Arc Length, Through-Wall Flaw . ... ... .. . ... ... .. . .. ... ... . .. ... ... ... .. . ... ... .. . .. ...... .. . . 11 3.3.3 Weld Metal .. . .. ... . .. . .. ... ... ... .. . .. ... ... .. . .. ... . .. . .. ... ... ... ... ... ... .. . ..... . .. . .. ... ... .. . . 11 3 .4 Conclusions - Structural Evaluation .. ... ... . .. ... ... .. . .. .... .. . .. ... ... ... ... ... ... .. . .. ... ... .. . . 11 3 .5 References . ... ... . .. ... ... ... .. . ... ... .. . .. ... ... .. . .. ... ... ... ... ... ... .. . ...... .. . .. .... .. ... ... ... .. . .. ... ... ..12 4.0 Method for Calculating Leakage. ... ... .. . .. . .. ... .. . .. ... ... . .. ... ... ... .. . ... ... .. . .. ... ... . .. ... ... .. . .. ... ... .. 19 4.1 Summary ... ... ... ... .. . ... ... .. . .. ...... .. . .. ... ... ... .. . ... ... .. . ........ . .. ... ... .. . ... ... .. . .. . ... .. ... ... ... .. 19 4.2 Modified B* Leakage Analysis. ........ . .. ... ... ... .. . ... ..... . .. ...... ... ... .. ... . .. ...... ... .. . .. ... 19 4.3 Calculation of Applicable Densities and Viscosity.. . .. ... ... .. . .. . ..... . .. ... ... ... .. . .. ... 25 4.4 Calculation of Limiting Leak Rates and Pressure Differentials ... . .. ... ... ... .. . .. ... 25 4.5 Calculation of Bounding Modified B* for Interim ARC Plants .. ... ... ........ .. . .. . 26 4.6 Conclusion .. . .. ...... ... ... ... .. . .. ... ... ... .. . .. ... ... . .. ... ..... . .. ... ... ... .. . ... ... .. . .. .... .. . .. ... ..... . .. . 27 4.7 References ... ... ... ... ... .. ... ... . .. ... ........ . .. ... ... .. . .. ... ... ... .. . ... ... .. . .. ... . .. . . . ... ... .. . .. ... ... ... .27 5.0 IARC Conclusions ... .. . .. .... .. . .. ... ... . .. ... .. ... ... ... .. ... ... . .. ... ... ... .. . ...... .. ... ...... .. . ... ... .. . .. ... ... ... . 35 5 .1 Limiting Structural Ligament ........ .. . ... ... .. . .. ... ... ... .. . .. ... ... . .. . . . ... . . . . . . . .... ... ... .. ... .. 35 5 .1 .1 Consideration of NDE Uncertainty ... ... ... .. . .. ... ... .. . ... ... .. . .. ... ... ... ... .. ... .. 35 5 .1 .2 Consideration of Crack Growth .. ... ... ........ . .. ... ... .. . .. ...... ... ... ... ... .. . .. ... .. 36 5.2 Leakage . .. . ... .. . .. ... ... . .. ... ........ . .. ..... ... ... ... ... ... .. . .. ... ... .. . .. ...... . .. ... ... ... ... ..... ... .. . ..... 36 5.3 References . . . . . . . . . . . ... .. . .. . .. ... ... .. . .. ... ... ... .. . .. ... ... ... .. ... ... . .. ... ... ... .. . .. ... . .. ... ... ... .. ... .. .37

LIST OF TABLES Table 3-1 Dimensions and Mechanical Properties of the Steam Generator Tubes ... ... ... .. . .. ... 13 Table 3-2 Interim Alternate Plugging Criterion Pressure Differentials . .. . .. ... ..... ... .. . .. . ... .. . .. ... 13 Table 3-3 Calculation of Required Minimum Ligament.. .... .. . .. ... ... . . . ... ... .. . .. ... ... ... .. . .. .... .. ... .. . 14 Table 4-1 List of H*/B* Plants . . . . .. ..... . .. ... ... ... ... .. . .. ... . .. . .. ... ... .. . ... ... .. . .. ... . .. . .. ... ..... . .. ... ... ... .. . .. . 28 Table 4-2 Primary to Secondary Leakage Data and Pressure Differentials for the Domestic Fleet. .. .. . ... ... .. . .. ... ... .. . .. . .. .... . . . .. . . . ... . .. ... ... .. . .. . ... .. . .. ... . .. . .. ... ..... . .. . .. ... ... .. . .. . 29 Table 4-3 Primary to Secondary Leakage Data and Pressure Differentials for the Domestic Fleet. .. .. . ... ... .. . .. ...... ... .. . .. .... .. . .. ... ... . .. ... ... .. . .. . ... .. . .. ...... . .. ... ..... . .. . .. ... . . . .. . .. . 29 Table 4-4 Primary to Secondary Leakage Data and Pressure Differentials for the Domestic Fleet. .. .. . ... ... .. . .. ... ... ... .. . .. .... .. . .. ... ... . .. ... ... .. . .. . ... .. . .. ... ... . .. ... .. ... . .. . .. ... ... .. . .. . 30 Table 4-5 Summary of Required Accident Length and Available Margin for the Domestic Fleet. .. .. . ..... ... . .. ... ... ... .. . .. .... .. . .. ... ... .. . .. . ... .. . .. ... . .. . .. ... ... . .. .. . ..... . .. ... ... ... .. . .. . 31 Table 5-1 PWSCC Growth Rates (Reference 3-6) . ... ... .. . ... ... ... ..... ... . .. .... .. ... ... ..... . .. . .. ... ... .. . .. . 38 Table 5-2 Calculation of Required Minimum Ligament for 18 and 36 Months Operating Periods ... ... ... ... ... .. . .. ... ... ... .. . ... .. . . . . . . ... . .. .. . ... ... .. . .. ...... . .. ... ... . .. .... .. ... .. . .. ... . .. ... .. .... .. . .. . 38

LIST OF FIGURES Figure 3-1 A Segment of a Steam Generator Tube Showing the Radial and Axial Axes as Well as the Crack Face . . .. .... .. . .. . ... .. . ... ... .. . ... ... .. . .. ... ... ... ... .. . ... .. . .. ... ... . .. ... . .. . .. ... . .. . .. . 15 Figure 3-2 The Geometry of a Partially Circumferential Crack on the Crack Surface Shown in Figure 3-1 .. ... . .. ...... . .. ... . . . . . .. ... .. . .. ... . .. . .. ... ... ... .. . .. ... . .. . .. ... ... .. . .. . ... .. . .. . ... .. . . 15 Figure 3-3 The Geometry of a Fully Circumferential Crack on the Crack Surface Shown in Figure 3-1 . .. . .. ... ... ... ... ... . .. .. . ... . .. ... ... . .. ... .. . . .. ... . . . ... .. . .. ... ... .. . ... ... .. . .. ... . .. . .. ... ... . .. ... . .. . . 16 Figure 3-4 A Schematic of a Conical Frustum Showing the Surface on Which the Crack Grows . ... ... .. . .. ... ... ... ... ... . .. ... ... . .. ... ... . .. ...... . .. ... ... ... .. . .. ... ... . .. ... ... .. . .. . ... .. . .. ... ... . .. ... ... .. 16 Figure 3-5 Schematic of a Partially Circumferential Flaw in the Weld Metal Along a Conical Frustum .. . .. ...... . .. ... ... ... .... .. . .. ... ... . .. ... ... ... .. . ... ... .. . .. ... ... .. . .. ...... . .. . ..... . .. .... .. . . 17 Figure 3-6 Schematic of a Fully Circumferential Flaw in the Weld Metal Along a Conical Frustum.. ... .. . .. ... ... ... ... . .. . .. ... ... ... ... ... ... ... ... . .. .. . ... ... .. . .. .... .. ... .. .... .. . .. ...... . .. ...... . .. ... ... .. 17 Figure 3-7 The Weld Metal Geometry and the Potential Crack Paths Considered . . ... .. . .. ... ... .. 18 Figure 3-8 A Schematic Representing an Infinitesimal Volume of Material in the Weld Metal Under the Applied Stress Tensor and Its Transformation to the Principal Stress Tensor . . ... .. . .. ... ... . .. ... ... . . . .. . . . . . . . . . . ... . .. ... ... ... .. . .. ... ... .. . .. .... .. . .. . .. ... . . . . . . . . . . . . .. . ... .. 18 Figure 4-1 Illustration of Tube-to-Tubesheet Crevice and Approximated Porous Medium Roughness of 125 pin is Typical of Installed Tube and Tubesheet Crevice Surfaces . ... .. . .. ... . .. . .. ... ... ... ... ... ... ... ... . .. ... ... ... .. . ... ... .. . . . ... ... .. . .. ... ... ... ... ... ... ... ... ... .. . .. ... 32 Figure 4-2 Plot of Loss Coefficient Data as a Function of Contact Pressure for Model F and Model D Steam Generators (Reference 4-2) .. . .. ... ... .. . .. ... ... ... .. . ... ... .. . .. ...... . .. ... 33 Figure 4-3 Plot of Loss Coefficient Data as a Function of Contact Pressure for Model 44F and Model 51 F Steam Generators (Reference 4-2) .. ........ . .. ... ... ... .. . ... .. ... . .. ... ... . .. ... 34 Figure 5-1 Correlation of Circumferential Crack Length and PDA ... . .. ... ... ... .. . ... .. ... . .. ... ... ... ... 39

1.0 INTRODUCTION

An alternate repair criterion (ARC) to limit the inspection depth in the tubesheet expansion zone, known as H*/B*, has been docketed by Wolf Creek Nuclear Operating Corporation since February 2006 and has been undergoing NRC review since that time . The H*/B* ARC seeks to minimize the depth of rotating coil inspection of the SG tubes within the tubesheet. The premise of H*/B* is that the expansion joint provides sufficient structural restraint to prevent the tube from pulling out of the tubesheet under normal operating and accident conditions, and that the accident induced leakage during accident conditions is bounded by a factor of two on the observed normal operating leakage. Because of the technical complexity of H*/B*, review of it cannot be completed in time for the Spring 2008 refueling outages.

This report provides technical justification for an interim alternate repair criterion (IARC) that requires full-length inspection of the tubes within the tubesheet but does not require plugging tubes if any circumferential cracking observed in the region greater than 17 inches from the top of the tubesheet (TTS) is less than a value sufficient to permit the remaining circumferential ligament to transmit the limiting axial loads (the greater of 3x NOP or 1 .4x SLB end cap loads) .

Axial cracks below 17 inches from the TTS are not relevant to the tube pullout arguments because axial cracks do not degrade the axial load carrying capability of the tube. Axial cracks do not require plugging if they are below 17 inches from the top of the tubesheet.

The calculation of the limiting circumferential ligament is provided in Section 3 of this report .

The calculation assumes that friction loads between the tube and tubesheet from any source are zero . This assumption avoids potential effects of uncertainties in tube and tubesheet material properties .

Also, based on the same assumption that the contact pressure between the tube and the tubesheet from any source is zero, this report provides a basis for demonstrating that the accident induced leakage will always meet the value assumed in the plant's safety analysis if the observed leakage during normal operating conditions is within its allowable limits . This analysis is provided in Section 4 of this report . The need to calculate leakage from individual cracks is avoided by the calculation of the ratio of accident induced leakage to normal operating leakage.

The tube-end weld is specifically excluded from the tube by TSTF-449, Rev. 4. Because friction between the tube and the tubesheet is ignored, the weld may become an important component in the transfer of the tube pullout loads to the tubesheet. Therefore, the minimum ligament necessary to transfer the pullout loads is also calculated in Section 3 . Because the tube-end weld is not considered a part of the tube, discussion of the inspection methodology is beyond the scope of this technical discussion. Discussion of how the weld will be examined is provided as a separate part of the license amendment request.

A bounding analysis approach is utilized for both the minimum ligament calculation and leakage ratio calculation. "Bounding" means that the most challenging conditions from the plants with hydraulically expanded Alloy 600TT tubing are used . Three different tube diameters are represented by the affected plants (11/16" dia., Model F; 3/4" dia. Model D5; 7/8"

dia., Model 44F) . The most limiting conditions for structural evaluation depend on tube geometry and applied normal operating loads . The conditions from the plant that result in the highest stress in the tube below the top of the tubesheet are used to define the minimum required circumferential ligament. The limiting leak rate ratio depends on the leak values assumed in the safety analysis and allowable normal operating leakage that results in the longest length of undegraded tube/crevice for assuring that acceptable leakage during the limiting design basis accident (i.e., steam line break, locked rotor and control rod ejection) above 17 inches below the tubesheet are used. The limiting cases for structural evaluation and leakage evaluation are not necessarily from the same plant. However, the resulting minimum ligament and required undegraded length of tube below the top of the tubesheet can be safely applied for any of the affected domestic plants identified in Table 4-l.

2 .0 PERFORMANCE CRITERIA The performance criteria of NEI 97-06, Rev. 2 (Reference 2-1) are the basis for these analyses .

The performance criteria are:

The structural integrity performance criterion is:

All in-service steam generator tubes shall retain structural integrity over the full range of'normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1 .4 against burst applied to the design basis accident primary-to-secondary pressure differentials . Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1 .0 on axial secondary loads.

The structural integrity performance criterion is based on ensuring that there is reasonable assurance that a steam generator tube will not burst during normal operation or postulated accident conditions .

The accident induced leakage performance criterion is :

The primary to secondary accident induced leakage rate for any design basis accident, other than a steam generator tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all steam generators and leakage rate for an individual steam generator. Leakage is not to exceed 1 gpm per steam generator, except for specific types of degradation at specific locations when implementing alternate repair criteria as documented in the Steam Generator Program technical specifications .

Primary-to-secondary leakage is a factor in the dose releases outside containment resulting from a limiting design basis accident . The potential primary-to-secondary leak rate during postulated design basis accidents shall not exceed the offsite radiological dose consequences required by 10 CFR Part 100 guidelines or the radiological consequences to control room personnel required by GDC-19, or other NRC-approved licensing basis.

The IARC for the tubesheet region is designed to meet these criteria. The structural criterion regarding tube burst is inherently satisfied because the constraint provided by the tubesheet to the tube prohibits burst. However, the structural integrity criterion is interpreted to mean

that tube pullout from the tubesheet is equivalent to a tube burst and must, therefore, be prevented.

The accident induced leakage criterion applies directly . The IARC will demonstrate that the accident induced leakage will not exceed the leakage assumed in the accident analysis for the plant which bounds all of the domestic plants which are anticipated to utilize the IARC .

2.1 REFERENCES

2-1 NEI 97-06, Rev.2, "Steam Generator Program Guidelines," Nuclear Energy Institute, Washington D.C., May 2005 .

3 .0 STRUCTURAL EVALUATION FOR MINIMUM CIRCUMFERENTIAL LIGAMENT

3.1 INTRODUCTION

An assessment to determine the remaining ligament in steam generator tubes (relevant to Model D, Model F, and Model 44F) necessary to support the assumed loading conditions in the presence of postulated, partially circumferential and fully circumferential flaws was performed.

Two locations were considered, within the steam generator tube wall at a location deep in the tubesheet and within the tube-to-tubesheet weld . In addition, growth of the crack was simulated by using four default primary water stress corrosion crack (PWSCC) growth rates.

Failure was determined to occur when the stress in the remaining ligament of tube or weld metal exceeded the flow stress .

3.2 ANALYSIS 3.2 .1 Description of the Steam Generator Models The tube geometries used in three models of steam generator which may utilize the IARC were analyzed. These were Model D, Model F, and Model 44F . The material properties applied in this analysis are LTL properties provided in References 3-1 through 3-4. The tube dimensions, material, and mechanical properties (at 650°F) are listed in Table 3-1 .

3.2.2 Flaw Geometries

1. Partial circumferentialflaw in the steam generator tube wall. This postulated flaw in the steam generator tube wall is assumed to have an initial depth of 0.010 inch and an initial arc length of 0.060 inch on the tube's inner diameter. The flaw extends from the tube's inner diameter to a depth of 0.010 inch such that the side faces of the flaw run parallel to the radii of the tube . Figure 3-1 shows a section of a steam generator tube, its radial and axial axes, and the crack face. Figure 3-2 shows the partial circumferential crack on the crack face. The initial depth and arc length are chosen to represent a typical surface flaw with a semi-elliptic shape and a 3 :1 aspect ratio subject to mode I crack opening (Reference 3-5). Thus, the length of the semi-major axis is initially three times that of the semi-minor axis, and the tensile axis of the load which opens the crack is normal to the direction of crack propagation. The initial depth of 0.010 inch is a commonly accepted initial flaw depth upon initiation. The flaw simultaneously grows by PWSCC both radially and circumferentially, and it maintains its initial shape. Upon breaching the outer diameter of the tube, the flaw continues to grow circumferentially until the remaining area of the tube cannot support the applied loading.

2. Full circumferential flaw in the steam generator tube wall. The postulated, full circumferential flaw in the steam generator tube wall is assumed to have an initial depth equal to 0.010 inch, consistent with the partial circumferential flaw . The

depth is also measured from the tube's inner diameter . Figure 3-3 shows the geometry for this type of flaw . This type of flaw grows by PWSCC radially only until the remaining ligament can no longer support the applied loading.

3. Partial circumferential, through-wall flaw in the steam generator tube wall. This type of geometry was chosen to correspond to the type of flaw that may exist upon detection. The assumed initial arc length of this flaw is 40 degrees, and the flaw grows by PWSCC circumferentially only until the remaining ligament can no longer support the applied loading. The geometry for this flaw is identical to the geometry shown in Figure 3-2 with the exception that the crack depth is through-wall.

4. Partial circumferential flaw in the weld metal. This geometry is similar to that described in number 1 above, except that it is in the weld and grows due to PWSCC in the shape of a conical frustum on an angle determined by the plane of maximum principal stress . The initial depth and arc length are 0.010 inch and 0.060 inch, respectively . Figure 3-4 is a schematic of a conical frustum and the surface on which the crack grows, and Figure 3-5 is a schematic of the flaw on that surface. The growth is simultaneously radial and circumferential until the remaining ligament cannot support the applied loading.

5. Full 360 degree circumferential flaw in the weld metal. This flaw, of 0.010 inch initial depth grows radially only due to PWSCC. It also grows in the shape of a conical frustum on an angle determined by the maximum principal stress until the remaining ligament cannot support the applied loading. Figure 3-6 is a schematic of this flaw geometry .

3.2 .3 Initiation Implicit in the preceding section is that the flaws are presumed to exist as the initial condition for the crack growth cycle. A crack growth cycle as defined in this analysis is full power operation for the length of time for the crack to grow from initial conditions until the minimum residual ligament is attained. The time variable is important to establish the ultimate required residual ligaments for different planned plant operating periods between inspections.

3.2 .4 Pressure Loading for Flaws in the Tube Wall The requirement for tube integrity is that the tube be able to support loads due to a pressure difference of 3*APNOP or 1 .4

• APSLB, whichever is more limiting. A review of the data available shows that the most limiting condition is due to APNOP of Surry Units 1 and 2 [

]a" Therefore, the most limiting pressure differential to determine end cap loads is based on 3 *1 APNOP of the Surry Units 1 and 2 and equals [ ]a,c,e This is conservative relative to the actual loads . Once a PWSCC flaw initiates, the faces of that flaw are subject to internal pressure, which in this case is the primary side pressure (2250 psia).

3.2.5 Pressure Loading Effects in the Weld Metal The plants being addressed for this study all have flush welds . The weld is assumed to have an elliptic shape with a semi-major axis equal to the tube wall thickness, a semi-minor axis equal to 0.014 inch, and a crown extending 0.008 inch below the tubesheet cladding surface. This is a conservative idealization of the actual weld nugget. In-process measurements of the welds have determined that the weld protrusion from the tubesheet surface is between 0.008 inch and 0.013 inch. Also, visual examination of the welds show that the autogenous weld nugget is elliptical and inclined to horizontal with the interface between the weld and the tube approximately 0.035 inch into the tubesheet bore. Therefore, the idealized representation of the weld is conservative to the actual manufacturing condition.

Three main crack paths are most likely to occur due to the applied loading. One is the horizontal surface between the tube bottom and the weld . In the most idealized fashion, the end cap loads result in a tensile stress along this interface. The second crack path is the vertical line from the tube-tubesheet interface to the bottom of the weld metal. In this case, the end cap loads result in a shear stress along this line of crack propagation. The third crack path is in the weld metal, between the previous two paths, and whose loading is a combination of tensile stress and shear stress . Figure 3-7 is a schematic of the weld geometry and the crack paths just discussed. The simplifying assumption used in this study is that the stress tensor of an infinitesimal volume of material in this region is comprised of the stress components calculated for the first two crack paths. This results in the maximum principal stress acting on a line that is approximately 35 degrees counter-clockwise from the tube bottom, where the center of rotation is 0.020 inch above the bottom surface of the tubesheet cladding and along the tube-tubesheet interface. Figure 3-8 is a representation of an infinitesimal volume of material, the applied stress tensor, and the principal stresses . As the crack grows, a decreasing area of the weld metal is subject to the maximum principal stress, however the flaw area is then subject to internal pressure on its faces.

3.2.6 Constraint The tube region subject to cracking is deep in the tubesheet (>17 inches below the top of the tubesheet) . The tubes are assumed to be flush against the tubesheet due to the hydraulic expansion process; however, there is no interference force due to pressure . No motion is possible in the lateral direction. Furthermore, it is also assumed that there is no friction acting on the joint between the tube and the tubesheet. The result of these assumptions is that only vertical displacement is allowed and the stresses in the tube wall are purely tensile; there is no bending stress component because of the lateral restraint of the tubesheet. Similarly, the weld metal is subject only to the tensile loads transmitted by the tube . Therefore, any crack in the weld metal will also open in a purely tensile mode . This is the reason that a weld crack in a direction radiating away from the tube's centerline is not considered here. In this case, the residual weld nugget on the tube results in mechanical interference with the residual weld nugget on the tubesheet, and the tube cannot pull out of the tubesheet.

3.2.7 Force Balance 1 . Partial circumferential flaw in the steam generator tube wall. The force balance for this scenario is one in which the end cap load plus the force due to the internal pressure acting on the faces of the flaw is balanced by the force reacted over the tube wall's cross-sectional area minus the flaw area. As the flaw grows, the areas of both the tube wall cross-section and the flaw change. The equation used in this part of the study is ax,e where

]a,c,e P is the pressure [

P; is the internal pressure (2250 psia),

r; is the inner radius of the steam generator tube, d is the crack depth, A 9 is the arc length of the crack, his the stress reacted by the steam generator tube's cross-section, and ro is the outer radius of the steam generator tube.

2. Fully circumferential flaw in the steam generator tube wall. The force balance dictated by this case is one in which the end cap load plus the internal pressure acting over the crack faces of a fully circumferential flaw is balanced by the force reacted by the steam generator tube wall's cross-sectional area minus the area of the flaw. Again, the areas of both the flaw and the steam generator tube wall's cross-section change as the flaw grows. The equation used to model this situation is C

where the variables are the same as previously defined .

3. Partially circumferential, through-wall flaw in the steam generator tube wall. This situation is identical to scenario 1 with the exception that the initial flaw is through-wall at the beginning of the crack growth cycle, and the initial arc length of the flaw is 40 degrees . This models a reasonable flaw length that would be detected by

+Pt inspection which is assumed to be throughwall . The force balance for this case is a,c,e

where the variables are the same as previously defined.

4. Partial circumferential flaw in the weld metal. The welds applicable to the plants under consideration are flush welds. Thus, the weld was modeled as an ellipse.

The starting point of the ellipse region is the steam generator tube wall's inner diameter. This case is one in which normal stress and shear stress components are present. The normal stress results from a potential crack propagation path that runs along the interface between the steam generator tube wall and the weld metal. The shear stress component is from a potential crack propagation path that runs vertically from the interface between the steam generator tube and the tubesheet to the crown of the weld . The infinitesimal element of weld metal is assumed to have the normal and shear stress components that result from each of the two crack propagation paths (assuming that only one is active and the other is fixed) . Hence, the normal stress component used is a,c,e and the shear stress component is a,c,e b is the semi-minor axis (0.014 inch). The three principal stresses that result from calculating the invariants of the stress tensor comprised of the above components are:

a,c,e and the direction of the principal axes is determined by:

]a,,,' extending The crack propagation direction is found to be approximately [

from the steam generator tube-tubesheet interface toward the centerline of the steam generator tube. This results in a crack propagation surface that is an inverted frustum of a cone. Using the surface of revolution technique (see Reference 3-6), the surface area of this conical frustum is a,c,e where e is the approximately [ angle defined above, y is the vertical

]a,e,a location of the intersection of the crack propagation line and the ellipse, and the rest of the variables are defined for scenario 1 above. The area of a flaw extending a depth d into this surface and over an arc length 0~ extending over this surface is a,c,e 7

where all of the variables have been previously defined. The resulting force balance for this scenario is C

where, in this case, a is the stress reacted by the remaining surface area of the frustum.

5. Full circumferential flaw in the weld metal. This number is similar to number 4 with the exception that the flaw is now fully circumferential . The area of the flaw in this case is a,c,e 7

The resulting force balance is a,c,e where, again, his the stress reacted by the remaining surface of the frustum.

3 .3 RESULTS AND DISCUSSION The required remaining ligaments are shown in Table 3-3 . The required remaining circumferential ligaments for initially non-360 degree throughwall circumferential flaws are expressed in terms of degrees of arc. The required remaining radial ligaments for full 360 degree non-throughwall circumferential flaws are expressed in terms of inches .

10

3.3.1 Steam Generator Tube Wall Cross-Section The values contained in Table 3-3 indicate that the required remaining ligament for partially

]a'e,e circumferential flaws is approximately [ while the required remaining ligament

]a,c,e for fully circumferential flaws is approximately [ The Model F steam generator tube requires less remaining ligament than do either the Model D or Model 44F steam generator tubes.

3.3.2 Steam Generator Tube Cross-Section with an Initial 40 Degree Are Length, Through-Wall Flaw The results contained in Table 3-3 show that a partially circumferential flaw that is initially through-wall requires about the same remaining ligament of material as the case for which the initial flaw was not initially through-wall Since the force balance is based on net tensile force, this result is expected.

3.3.3 Weld Metal The results for the weld metal calculations are also shown on Table 3-3 . The required remaining ligaments for both the partially circumferential and fully circumferential flaws are approximately [ .c,` ]a,e,e

]a arc length and approximately [ for the partially circumferential and fully circumferential flaws, respectively , significantly less than required for the steam generator tube wall.

This situation for the weld is mechanically different than for the steam generator tube wall. In the latter case, the pressure differential that causes the end cap load is based on the internal pressure which acts on the flaw's faces. The end cap loading relieved in the wall during crack growth is replaced by another pressure loading on the crack faces. For the weld, the pressure differential causes an end cap load, which in turn results in a maximum principal stress along an inclined crack propagation path . The maximum principal stress [ ]a,c,e is much greater than the initial stress reacted by the steam generator tube wall [ ]a,c,e However, as the flaw grows in the weld metal, it is the maximum principal stress in the area of the flaw that is relieved and replaced with the primary pressure loading [ ]a,,,, over the crack faces. In addition, the surface area relevant to the weld metal is slightly larger than that contained in the steam generator tube wall due to its incline.

3.4 CONCLUSION

S - STRUCTURAL EVALUATION

" The required arc of ligament for an initial, partially circumferential flaw of 0.010" depth in the steam generator tube is approximately [ ]a,c,e In general, the Model F steam generator tube wall requires the least amount of remaining ligament.

However, Model F requires the least amount of time to grow to its critical flaw size .

The results of all of the calculations performed are enveloped by an arc length of ligament equal to [ ]" for this geometry .

" The required arc of ligament for the case when the initial flaw is through-wall over a 40 degree arc is approximately the same as above. This is expected as the critical flaw size is based on net tensile stress . An arc length of ligament equal to [ ]a,e,e is necessary to bound the results for this geometry .

" Initial, fully circumferential flaws in the steam generator tube can grow to approximately through-wall before failure was calculated to occur. The

]a'c'e minimum required radial ligament depth is [ for the bounding case . This is provided for information only since the underlying assumption of the IARC is that circumferential cracks will be considered 100% throughwall .

Ia'c,e

" Initial, partially circumferential flaws in the weld required a [ arc of remaining weld material, significantly less than the arc required in the steam generator tube wall. In order to bound the results for this geometry, an arc length of material

,,,e spanning [ ]a is required.

" Initial, fully circumferential flaws in the weld metal were able to grow to approximately [ ]a,e'e through-wall before failure was calculated to occur, again significantly less than the ligament required in the steam generator tube wall . A bounding value of [ Ia,c,e of ligament is required for this case . This is provided for information only since the underlying assumption of the IARC is that circumferential cracks will be considered 100% throughwall.

3.5 REFERENCES

3-1 J. A. Begley and J. L. Houtman, WCAP-12522, "Inconel Alloy 600 Tubing-Material Burst and Strength Properties," January 1990.

3-2 G. Whiteman, WCAP-16670-P, "Steam Generator Alternate Repair Criteria for Tube Portion within the Tubesheet at Comanche Peak Unit 2," November 2006 .

3-3 WCAP-16124-P, "Justification for the Partial-Length Rotating Pancake Coil (RPC) Inspection of the Tube Joints of the Wolf Creek Model F Steam Generators," September 2003 .

3-4 G. Whiteman, WCAP-16506-P, "Steam Generator Alternate Repair Criteria for Tube Portion within the Tubesheet at Turkey Point Units 3 and 4," December 2005 .

3-5 T. L. Anderson, Fracture Mechanics Fundamentals and Applications , Second Edition, New York : CRC Press, 1995 .

3-6 G. B. Thomas, Jr., Calculus and Analytic Geometry, Alternate Edition, Reading: Addison-Wesley Publishing Company, Inc., 1972 .

12

Table 3-1 Dimensions and Mechanical Properties of the Steam Generator Tubes Model D F 131 44F 141 O.D. in 0.764 0.703 0.893 Wall Thickness 0.04257 0.0396 0.0495 in I.D. in 0.664 0.6075 0.775 Material Alloy 600 Alloy 600 Alloy 600 Heat Treatment Thermally Treated Thermally Treated Thermally Treated a,c,e ro in 0 .382 0.3515 0.4465 r; in 0.33943 0.3119 0.397 Note [1] : These properties listed are lower tolerance limit (LTL) properties from Reference (3-1).

Note [2]: The expanded tube outer diameter and thinned wall dimensions for the Model D steam generator tubes are from Reference (3-2).

Note [3]: The expanded tube outer diameter and thinned wall dimensions for the Model F steam generator tubes are from Reference (3-3).

Note [4]: The expanded tube outer diameter and thinned wall dimensions for the Model 44F steam generator tubes are from Reference (3-4).

Table 3-2 Interim Alternate Plugging Criterion Pressure Differentials Plant Normal Steam Source Document Operation Line OP (psi) Break OP si a,c,e 13

Table 3-3 Calculation of Required Minimum Ligament Circumferential Minimum Extent of Flaw Structural Ligament a,c,e

Figure 3-1 A Segment of a Steam Generator Tube Showing the Radial and Axial Axes as Well as the Crack Face a,c,e Figure 3-2 The Geometry of a Partially Circumferential Crack on the Crack Surface Shown in Figure 3-1 15

Figure 3-3 The Geometry of a Fully Circumferential Crack on the Crack Surface Shown in Figure 3-1 Figure 3-4 A Schematic of a Conical Frustum Showing the Surface on Which the Crack Grows 16

Figure 3-5 Schematic of a Partially Circumferential Flaw in the Weld Metal Along a Conical Frustum a,c,e Figure 3-6 Schematic of a Fully Circumferential Flaw in the Weld Metal Along a Conical Frustum 17

Figure 3-7 The Weld Metal Geometry and the Potential Crack Paths Considered Figure 3-8 A Schematic Representing an Infinitesimal Volume of Material in the Weld Metal Under the Applied Stress Tensor and Its Transformation to the Principal Stress Tensor .

(This element is in the weld metal to the left of the shear plane vertical line in Figure 3-7.)

18

4 .0 METHOD FOR CALCULATING LEAKAGE 4.1

SUMMARY

The alternate repair criterion (ARC) known as B* (Reference 4-2, 4-3), for "bellwether" approach, specifies the length of sound tubing required for the tube portion within the tubesheet that will assure that a plant's accident induced primary-to-secondary (P/S) leakage limit will not increase greater than a factor of two (2) above the normal operating leakage. The B*

criterion relies on the contact pressure between the tube and the tubesheet . Technical issues remain to be resolved in the calculation of contact pressure between the tube and the tubesheet.

Therefore, a modified B* approach is presented in this section which demonstrates that a plant with postulated cracks in the tube portion within the lower four inches of the tubesheet will still meet the accident induced leakage limits for safe steam generator operation under the assumption that no contact pressure exists between the tube and the tubesheet.

The modified B* approach shows that for an undegraded 17 inch depth of tube, measured from the secondary side surface of the tubesheet, there is a margin of a factor of 1 .7 on the limiting length below the neutral axis of the tubesheet required to meet accident induced leakage limits for the bounding plant among those under consideration. This result means that, for the bounding plant, a 17 inch length of tube in undegraded condition provides more than 1 .7 times the length of porous medium (crevice) necessary below the neutral axis of the tubesheet to limit the accident induced leakage to the value assumed in the safety analysis.

Figure 4-1 shows a sketch of the porous medium in the tube-to-tubesheet crevice. The typical machining finish of 125 micro-inches defines the porosity, but is assumed to provide no interlocking or friction.

A summary of the plants that are included in the modified B* analysis is given in Table 4-5.

Based on the plant information, the ratio of the allowable accident leak rate to the allowable normal operating leakage limit in the bounding case steam generator is two (2). This value ranges from two (2) to six (6) for the plants under consideration for the IARC . See Table 4-2.

This means that the leakage during accident conditions can increase by no more than 2 to 6 times the leak rate during normal operating conditions for the plants under consideration. This section shows that ample margin exists in undegraded crevice length for the bounding plant.

The results for the bounding plant envelope all of the plants under consideration.

4.2 MODIFIED B* LEAKAGE ANALYSIS The approach to the modified B* leakage analysis is similar to that used in the original B*

(Reference 4-2) . Where B* calculates the length of undegraded tubing, measured from the TTS, required to equilibrate the flow resistance during normal operating and during accident conditions so that the increase in primary to secondary leakage is limited to a function of the ratio of the pressure differential during the limiting design basis accident and normal operating conditions, the Modified B* analysis calculates the ratio of undegraded crevice length determined by eddy current inspection to the length of undegraded crevice required to meet the design basis accident analysis primary to secondary leakage assumption . By definition of the 19

IARC, 17 inches from the TTS is the available undegraded crevice length because confirmed cracking in this length will require the tube to be plugged. Both the pressure difference ratio and the ratio of the length of crevice during normal operating and the limiting design basis accident are factored into the margin determination as discussed below. By definition, the plant with the smallest allowable accident analysis leakage assumption results in the longest crevice length necessary to assure that accident analysis leakage assumptions are not exceeded. For the plants in question, the Modified B* value ranges from a safety factor of [ down to

]a'c'e

[

]" at a distance 17 inches below the top of the tubesheet (See the "n" values in Table 4-5).

Conservatively using the neutral axis as a reference point, the Modified B* value ranges from

]a

[ ]" down to [ '°'e (See the "n"' values in Table 4-5). Again, these values are the ratio of undegraded tube/crevice length confirmed by eddy current inspection to the length of undegraded crevice calculated using the D'Arcy equation necessary to preclude exceeding the limiting design basis accident analysis leakage assumption .

The D'Arcy formula for axial flow in a porous medium is used to calculate the leakage ratio and to evaluate the potential resistance to leakage in the crevice of the tubesheet . Other available leakage models (Bernoulli, Orifice Flow) are known to be less conservative than the D'Arcy model. Unresolved technical issues regarding the calculation of contact pressure between the tube and the tubesheet in the original B* require that both the bellwether principle and the application of D'Arcy's law do not employ contact pressure equations or relationships in the leakage analysis .

The D'Arcy model for describing axial flow in a porous medium, taken from Reference 4-1 is:

Where:

Q is the flow rate for the fluid through the medium, Ap is difference in pressure (or driving head) acting to force the fluid through the medium, is the viscosity of the fluid, K is the resistance to flow through the medium and 1 is the axial length of the medium.

The term pKl is the flow resistance, R. In that case, (1) becomes

which produces a relationship between fluid flow, flow resistance and driving potential similar to electrical currents (i.e., I = Y/R) and allows for similar analogies and assumptions to be made . See Figure 4-1 for a sketch of the system used to describe the porous medium present in the annulus of the tubesheet crevice.

In the following discussion the term R' refers to uK and the axial length of the porous medium is left in the equation as a separate variable as shown in Equation (3).

Note that in previous submittals (Reference 4-2, 4-3), the length of the medium was included in the term R (see equation 2), which led to the conclusion that if the resistance of the crack and tubesheet crevice to leakage during normal operating (NOP) conditions was equal to the resistance of the crack and tubesheet crevice during steam line break (SLB), the increase in leakage between NOP and SLB conditions would be governed solely by the pressure differential. The original bellwether ratio of the expected accident leak rate to the required normal operating leak rate of 2 was based on this assumption because the pressure differential at SLB conditions is approximately double that during normal operating conditions. Therefore, the leakage during SLB conditions would be limited to twice that of the leakage during NOP for a length of crevice and a location of the leak that validates the assumption of equal resistance between SLB and NOP conditions.

The purpose of the interim ARC leakage assessment is to calculate the length of porous medium (crevice) required to limit primary-to-secondary (P/S) leakage to an acceptable level during a postulated SLB (or limiting design basis accident) to provide adequate resistance and margin against leakage during accident conditions assuming no contact pressure between the tube and the tubesheet exists . This length is defined as Modified B* and is used to assess the potential for leakage and acceptability of leakage flow rates assuming a full depth inspection of the tube portion with the tubesheet and a 17" length of tube free of all cracking indications .

The Modified B* ratio is prescribed as the accident analysis limit divided by the plant Technical Specification limit of 0.1 gpm.

The margin against leakage during an accident event can be defined using equations (1) and (3).

An example calculation of the modified B* ratio and the required length of porous medium necessary to accommodate the limiting accident leakage is provided below for the limiting case of zero contact pressure . There is no contact pressure between the tube and the tubesheet (P,o c~,t = 0 psi) but the tube and the tubesheet are assumed to remain in contact. Assume that a point exists where the viscosity and leakage resistance during normal operating conditions will be equal to that of the viscosity and leakage resistance during accident conditions at some elevation in the tube-to-tubesheet crevice. That is, R' NOP = R' DBA =R'

In this case the resistance to flow is calculated assuming that the liquid must flow through a tortuous path that begins at the crack (primary side) and ends at the top of the tubesheet (secondary side). No credit is taken for the increase in contact pressure between the tube and the tubesheet due to tubesheet flexure during accident conditions which would increase the resistance to flow through the crack and crevice.

The following example demonstrates the approach :

If the limiting leakage during NOP is 0.1 gpm and the leakage assumed in the safety analysis for SLB is 0.35 gpm, the ratio between SLB and NOP leakage is:

QSLB = 0 .35 = 3,5 QNOP 0 .10 Note that prior knowledge of the shape or orientation of the flaws that contribute to this leakage is not required . The ratio merely reflects the total leakage volume to which the plant is limited during operation. The ratio of the leak rates can be calculated using equations (3) and (4) which gives QSLB _ APSLB R' NOP GOP QNOP OPNop R' SLB 1SLB

_R'1 QSLB _ OPSLB NOP _ OPSLB _lNOP QNOP APNOP R' IsLB APNOP 1SLB QSLB S_LB I NOP QNOP Ap NOP ISLB Substitution of the pressure differentials and the limiting leak rate ratio into equation (5) yields the ratio of the porous medium (crevice) length necessary to maintain the limiting accident analysis leakage assumption . For example, if the limiting primary to secondary pressure differential during normal operations is 1274 psig and the limiting accident pressure differential is 2560 psig the required length ratio for a leak ratio of 3.5 is given by :

2560 INOP 3 .5 =

1274 ISLE 1 1VOP 1274 3 .5 3 .5 -1 .74 ISLB 2560 2.009 I NOP =1 .74 ISLB 22

The length ratio can be used with the data for loss coefficient and viscosity to calculate the required length of tube and crevice necessary to match the limiting leakage flow rate. If the leakage limits for the operating SG are based on "hot" or operational conditions, then the viscosity of the single phase leaked fluid is approximately equal to the viscosity of liquid water at 600°F .2 The viscosity of liquid phase water at 600°F is approximately 1 .76E-6 lbf-s/in2 (Reference 4-2) . The loss coefficient data given in WCAP-16794-P (Reference 4-2) shows that for a contact pressure of approximately 0 psi, the bounding loss coefficient from the 95%

]a" confidence interval fit is equal to [ The value of loss coefficient that

]a,e,e approximately bounds all of the test data is [ (See Figures 4-2 and 4-3).

Note that the primary to secondary leakage at 600°F that corresponds to 0.1 gpm at room temperature conditions is 0.14 gpm. It is necessary to adjust the limiting leak rate for the NOP conditions because the loss coefficient data in WCAP-16794-P (Reference 4-2) is adjusted to represent room temperature conditions . Using the bounding loss coefficient value and the viscosity to calculate the required length of porous medium (crevice) to accommodate the NOP leakage gives Q=

pKI

_ OPNOP INOP PNOPKQNOP a,c,e 76440.00 =1 INOP = .340 56918.40 Recall that:

INOP =1 .74 ISLB Therefore, the length of tube and crevice necessary to maintain the limiting leakage flow rate at accident conditions is Modified B* =1SLs =1.34 /1.74 = 0.77in 2 : The viscosity and loss coefficient are calculated at normal operating conditions because the normal operating conditions for the set of plants seeking to use the IARC are more closely related . Also, it is conservative to assume that the viscosity of the liquid phase of water during SLB equals the viscosity of the liquid phase of water at NOP condition .

23

This result shows that the length of porous medium required during the normal operating condition is more limiting compared to the length of porous medium required during an accident condition.

Inspection of the tube to a depth of 17 inches to ensure that the tube is free of cracking indications means that there is at least 17 inches of tube material and crevice to interact and provide leak resistance . Therefore, the available factor of safety against leakage in excess of accident analysis assumptions, n, is 17 n= z- 22 0.77 The result for n shows that there is greater than a factor twenty (20) times the length of tube and crevice annulus/porous medium necessary to maintain the maximum allowable leakage limits for plant operation during steam line break conditions in this example.

It is possible for the tubesheet to deflect during operations as the pressure differential from the primary to secondary surface varies so that the tubesheet crevices expand above the tubesheet neutral axis. It is reasonable to expect that the flow resistance of the crevice will decrease as the tubesheet crevice expands. The tubesheet deflection will tend to expand the crevice from the neutral axis of the tubesheet to the secondary side face of the tubesheet in the near and mid-range radii. In the context of this analysis the term near radius refers to the tubesheet radii from the center to a distance of 20 inches, mid range refers to the radius from 20 inches to 40 inches and peripheral refers to tubesheet radii greater than 40 inches from the center. The tubesheet deflection will tend to constrict the tubesheet crevice from the neutral axis to the primary face of the tubesheet in the near and mid-range radii. The effects of the tubesheet deflection are reversed in the peripheral radii so that the crevice tightens above the neutral axis and expands below the neutral axis . In order to accommodate this phenomenon, the available tube-to-tubesheet crevice or available porous medium is only that length within the tubesheet, above or below the neutral axis, which experiences constriction of the tubesheet bore . This will be the reference available crevice length in this analysis. This means that even though there are 17 inches of undegraded crevice available due to the IARC assumptions, only that difference between the neutral axis and 17 inches is assumed to act to provide leakage resistance . In the case of a Model F steam generator the neutral axis is located approximately [ ]a°,e below the secondary side face of the tubesheet (Reference 4-2). This means that for a Model F steam generator there is a [ ]"e,e long length of porous medium available to resist leakage that can be assured to not dilate due to tubesheet flexure. Following the example above this means that the actual factor of safety against exceeding the accident induced leakage is:

a,c,e This result for n' indicate that if the region of the tubesheet crevice affected by rubesheet bow is removed from consideration there is at least a factor of eight (8) on the available porous medium to resist accident and normal operating leakage in this example.

24

4.3 CALCULATION OF APPLICABLE DENSITIES AND VISCOSITY Calculation of the leaked fluid density and the applicable viscosity during NOP conditions is required to determine the required length of porous medium. The density of the leaked fluid is important because different operating plants use different leakage assumptions in their safety analyses . For example, a plant may assume that the leaked fluid is "hot" or at operating temperature, which means that the volume of the fluid is increased relative to a "cold" or room temperature condition. Some of the potential plants under consideration have revised the Plant Technical Specifications to use a mass flow rate for the leakage limit which removes the concern of "hot" or "cold" volumes entirely. The modified B* analysis assumes that all leakage volumes are "cold" leakage volumes even though some plant values for accident analysis leakage are at operating conditions. This results in a lower ratio value for allowable leakage rate during design basis accident conditions to normal operating leakage limit and longer required crevice lengths during the design basis accident .

The modified B* analysis also assumes that the fluid viscosity during NOP bounds the viscosity during any accident at lower temperatures . The viscosity term appears in the denominator of equation (3) so it is conservative to keep it at a lower value which reduces the denominator (viscosity of water increases at lower temperatures) and increases the required length of porous medium.

4.4 CALCULATION OF LIMITING LEAK RATES AND PRESSURE DIFFERENTIALS The Modified B* IARC leakage analysis represents a bounding approach that describes the limiting leak and length ratios for the potential user plants that are noted on Figure 4-1 . These plants meet the definition of an H*/B* plant; that is, steam generators with Alloy 600TT tubing that is hydraulically expanded over the full depth of the tubesheet .

The limiting leak rate ratio, accident induced leakage to normal operating leakage, for the plants on this list is the lowest leak rate ratio for any plant, which is two (2). The bounding analysis for the modified B* must justify a leak rate ratio of two (2) . The limiting leak rate ratio is taken from Catawba Unit 2 and is assumed to be a cold volume . No leak rate ratio higher than six has been identified (See Table 4-2).

Table 4-2 through Table 4-5 show the accident and normal operating condition leak rates and the associated pressure differentials for each condition. The pressure differentials are calculated assuming hot leg, low TAVG properties for NOP conditions.

The inputs for the calculation of the limiting length of porous medium (crevice) and the limiting leakage ratio are applied consistently . That is, the pressure differential and leak limit for a single plant is used to calculate the porous medium length and the available margin at 17 inches . The longest required length that bounds all of the other plants under consideration is then taken as the bounding, or limiting length, for all of the plants .

4.5 CALCULATION OF BOUNDING MODIFIED B* FOR INTERIM ARC PLANTS Applying the limiting leak rate and pressure differential data from Table 4-2 in Equation (5) gives a length ratio of [ ]a,,,, . The calculation of the limiting length ratio is given below QSLB __ APSLB _ 1 NOP QNOP '4NOP 1SLB a,c,e a,c,e a,c,e Calculating the required length of porous medium (crevice) for the limiting plant during NOP conditions yields

_ OP NOP INOP NOP kQNOP a,c,e a,c,e C

a,c,e

Therefore, the 17 inch length of undegraded crevice within the tubesheet provides more than

[ ]"e times the required length required to meet the accident induced leakage limits for the bounding plant. The [ ]a'e'a inch length of undegraded tubing below the neutral axis provides more than [ f" times the required length of crevice required to meet the accident induced leakage limits for the bounding plant. The result for the bounding plant envelopes all of the other plants under consideration (see Table 4-5) and the margin for all other plants in Table 4-5 is greater. Therefore, the limiting modified B* result of [ ]a'c'e inches is a bounding result for all of the plants under consideration .

4.6 CONCLUSION

A basis is provided to assure that the accident induced leakage for the limiting accident will not exceed the value assumed in the safety analysis for the plant.

The length of undegraded crevice required to limit the accident induced leakage to less than the value assumed in the safety analysis for the limiting plant is [ ]a'c'e inches . By definition of the IARC, a tube that can remain in service has an undegraded crevice of 17 inches . Therefore,

],°`,e a factor of safety of [ is available. Expressed in length terms, the length margin in the crevice is [ ]a,c,e inches .

For all IARC candidate plants other than the limiting plant, the margins on length required to limit the accident induced leakage to less than the value assumed in the safety analysis is greater.

In summary, no leakage issue is associated with the IARC unless the normal operating leakage attributable to the tubesheet expansion zone (TEZ) is greater than its limit. Continued operation of the plant with leakage greater than the specified allowable limit is not possible .

4 .7 REFERENCES 4-1 . NSD-RMW-91-026, M.J. Sredzienski, "An Analytical Model for Flow Through an Axial Crack in Series with a Denting Corrosion Medium."

02/05/1991 .

4-2 . WCAP-16794-P, G.W. Whiteman, "Steam Generator Tube Alternate Repair Criteria for the Portion of the Tube Within the Tubesheet at the Vogtle 1 & 2 Electric Generating Plants ." 10/2007.

4-3 . Wolf Creek ET 07-0043; Docket No. 50-482 : "Response to Request for Additional Information Related to License amendment Request to Revise Steam Generator Program"; September 27, 2007 .

Table 4-1 List of H*/B* Plants Site a,c,e Alpha SG Model

Table 4-2 Primary to Secondary Leakage Data and Pressure Differentials for the Domestic Fleet .

SLB = Steam Line Break. LR=Locked Rotor. CRE=Control Rod Ejection.

NOP=Normal Operating Condition.

Plant Name Pressure (psi) P/S Le akage (GPM) LNop/LsLB LNOP LSLB a,c,e SLB NOP SLB NOP Ratio.

Table 4-3 Primary to Secondary Leakage Data and Pressure Differentials for the Domestic Fleet.

SLB = Steam Line Break. LR=Locked Rotor. CRE=Control Rod Ejection.

NOP=Normal Operating Condition .

Plant Name Pressure (psi)

LR NOP P/S Leakage (GPM)

LR . [ NOP Ratio I LNOP/LLR LNOP in LLR in a,c,e 29

Table 4-4 Primary to Secondary Leakage Data and Pressure Differentials for the Domestic Fleet.

SLB = Steam Line Break. LR=Locked Rotor. CRE=Control Rod Ejection.

NOP=Normal Operating Condition.

Plant Name Pressure (Tsi) P/S Leakage (GPM) LNop/LCRE LNOP LCRE a,c,e CRE NOP CRE NOP Ratio in in

Table 4-5 Summary of Required Accident Length and Available Margin for the Domestic Fleet.

Modified B*

Safety Margin Ratio a,c,e Plant Name LSLB LLR LCRE MAX in in in in l 7/LACCIDENT 6.5/LACCIDENT

Figure 4-1 Illustration of Tube-to-Tubesheet Crevice and Approximated Porous Medium Roughness of 125 pin is Typical of Installed Tube and Tubesheet Crevice Surfaces

Figure 4-2 Plot of Loss Coefficient Data as a Function of Contact Pressure for Model F and Model D Steam Generators (Reference 4-2)

Figure 4-3 Plot of Loss Coefficient Data as a Function of Contact Pressure for Model 44F and Model 51F Steam Generators (Reference 4-2)

5.0 IARC CONCLUSIONS 5.1 LIMITING STRUCTURAL LIGAMENT From Section 3 of this report, the bounding structural ligament required for the tube to transmit the operational loads is 115 degree arc. This assumes that the residual ligament is 100% of the tube wall in depth. For the tube-end weld, the bounding circumferential structural ligament is 35 degrees arc. A small circumferential initiating crack is predicted to grow to a throughwall condition before it is predicted to reach a limiting residual ligament . A residual ligament in a part-throughwall condition is not a significant concern, because of the assumption that all circumferential cracks detected are 100% throughwall.

5.1.1 Consideration of NDE Uncertainty The NDE uncertainty must be addressed to assure that the as-indicated circumferential arc of the reported crack is a reliable estimate of the actual crack. ETSS 20510.1 (Reference 6-1) describes the qualified technique used to detect circumferential PWSCC in the expansion transitions and in the TEZ. This technique is also considered qualified by the industry, and has been routinely used, for the detection of circumferential indications in the tack expansion region just above the tube-end weld . The qualification data is provided in the ETSS .

The fundamental assumption for the IARC is that all circumferential cracks detected are 100%

throughwall. Thus, even a shallow crack of small length will be considered to be throughwall.

Further, tube burst is not an issue for the IARC because of the constraint provided by the tubesheet; rather, it is axial separation of the tube that is the principal concern. Assuming that all circumferential cracks are throughwall reduces the inspection uncertainty to length of the cracks only . Further, the accuracy of the length determination is an issue only when the indicated crack approaches the allowable crack length (the complement of the required residual ligament) and if the indicated crack length is a reasonable estimate of the structural condition of the tube .

Prior investigations have correlated the axial strength of the tube to the Percent Degraded Area (PDA) of the flaw (Reference 6-2). PDA takes into account the profile of the existing crack, including non-throughwall portions and shallow tails of the crack. Using the data from ETSS 20510 .1 for cracks with a 90%, or greater, throughwall condition from both NDE and destructive examination, Figure 6-1 compares the actual crack length and corresponding PDA for the cracks to a theoretical PDA which assumes that all cracks are 100% throughwall. For all flaws greater than 60 degrees circumferential extent, the theoretical PDA line is bounding .

As the crack lengths increase, the separation of the actual PDA from the theoretical PDA tends to increase.

It is concluded that if the detected circumferential cracks are assumed to be 100% throughwall, the as-indicated crack lengths will be inherently conservative with respect to the structural adequacy of the remaining ligament. Therefore, no additional uncertainty factor is necessary to be applied to the as-measured circumferential extent of the cracks .

35

5.1 .2 Consideration of Crack Growth The growth of cracks due to PWSCC in the present study is dictated by four default PWSCC growth rates from Reference 6-3 . The distribution of growth rates is assumed to be lognormal.

Typical values and conservative values are given, although it is recommended in Reference 6-3 to use the default values only when the historical information is not available and not to use the typical values unless the degradation is mild. (No significant crack growth data exits for circumferential cracking in the tubesheet expansion region .) Both growth sets provided in Reference 6-3 have mean values and 95% upper bound values . See Table 6-1 . For this analysis, the typical 95% upper bound growth rate is used.

The residual structural ligament must be adjusted for growth during the anticipated operating period between the current and the next planned inspection. Typically, the operating periods for the affected plants are 18 calendar months ; however, some plants have planned outages in which no primary side inspections will be performed. Therefore, the cycle length adjustments are made to the minimum structural ligament required .

The circumferential growth rates are expressed as inches per EFPY in Table 6-2. Referring to Table 6-2, the maximum allowable throughwall circumferential crack size in a steam generator tube is 214° (=360° - 146° [required minimum ligament]) supporting one cycle of operation.

The maximum allowable circumferential crack size in a tube-to-tubesheet weld is 294° (360° -

66° [required minimum ligament]) supporting one cycle of operation.

5.2 LEAKAGE A basis, using the D'Arcy formula for flow through a porous medium, is provided to assure that the accident induced leakage for the limiting accident will not exceed the value assumed in the safety analysis for the plant if the observed leakage during normal operation is within its limits for the bounding plant. The bounding plant envelopes all other plants who are candidates for applying H*/B*. The D'Arcy formulation was previously compared to other potential models such as the Bernoulli equation or orifice flow formulation and was found to provide the most conservative results.

The length of undegraded crevice required to limit the accident induced leakage to less than the

]a,c'e value assumed in the safety analysis for the limiting plant is [ By definition of the IARC, a tube that can remain in service has an undegraded crevice of 17 inches . Therefore, a factor of safety of [ ]a°`°, is available. Expressed in length terms, the length margin in the

]a,c,e crevice is [

Significant margin on crevice length is available even if only the distance below the neutral axis of the tubesheet is considered . This distance is approximately [ ]"" During normal operating conditions, the tubesheet flexes due to differential pressure loads, causing the tubesheet holes above the neutral axis to dilate, and below the neutral axis, to constrict. No mechanical benefit is assumed in the analysis due to tubesheet bore constriction below the neutral axis of the tubesheet; however, first principles dictate that the tubesheet bore and crevice must decrease . Therefore, the leakage analysis provided is conservative .

36

For all IARC candidate plants other than the limiting plant, the margin on length required to limit the accident induced leakage to less than the value assumed in the safety analysis is greater than the values noted above for the bounding plant.

It is also concluded that if the normal operating leakage is within its allowable value, the accident induced leakage will also be within the value assumed in the bounding plants' safety analysis. This conclusion applies for all other plants which would benefit from implementation of the IARC .

5.3 REFERENCES

5-1 ETSS #20510 .1 ; Technique for Detection of Circumferential PWSCC at Expansion Transitions.

5-2 EPRI TR-107197; Depth Based Structural Analysis Methods for Steam Generator Circumferential Indications; November 1997 .

5-3 EPRI Document 1012987, "Steam Generator Integrity Assessment Guidelines, Revision 2," July 2006 .

Table 5-1 PWSCC Growth Rates (Reference 3-6)

Radial Circumferential Growth Direction

(%TW/EFPY in/EFPY Moan 4.5 0.04 Typical Values 95% U ner Bound 13 .1 0.12 Mean 7.0 0.08 Conservative Values 95% Upper Bound 20.4 0.24 Table 5-2 Calculation of Required Minimum Ligament for 18 and 36 Months Operating Periods Bounding EFPY Growth Growth Growth for Minimum Structural Required Structural (1) (In./EFPY) (Deg./EFPY) Operating Ligament Minimum Ligament (2) (3) Period (degrees) Ligament (degrees) (degrees)

Tube 18 CM Operation 1 .5 .12 20.65 31 F 146 36 CM 3 .0 .12 20.65 62 177 Operation 18 CM 1 .5 .12 20 .65 31 66 Operation Weld 36 CM 3 .0 .12 20 .65 62 L J 97 ejeration Notes:

4. It is conservatively assumed that 1 EFPY = 1 Calendar Year.

5 . 95% upper value of typical growth rates from Reference 6-3 .

6. Based on smallest Model F mean tubesheet bore dimension .

PDA vs TW Circ Crack Length TW -Theoretical

~. NDE>90%TW m NDE vs PDArnet o MET>90%

0 30 60 90 120 150 180 210 240 270 300 330 360 Circ Crack Length (degrees)

Figure 5-1 Correlation of Circumferential Crack Length and PDA

Attachment 6 Application for Steam Generator Tube Alternate Repair Criteria Technical Specification Amendment Braidwood Station, Units 1 and 2 Facility Operating License Nos . NPF-72 and NPF-77 NRC Docket Nos . 50-456 and 50-457 Westinghouse LTR-CDME-08-25 "Errata for LTR-CDME-08-11 ; "Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone"

Westinghouse Proprietary Class 2 Westinghouse To: P. J. McDonough Date: February 12, 2008 D. Alexander E. Arnold G.W . Whiteman J. A. Gresham E. P. Morgan From : H.O. Lagally Your ref :

Ext: 724-722-5082 Our ref : LTR-CDME-08-25 Fax: 724-722-5909

Subject:

Errata for LTR-CDME-08-I1 ; "Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone" The subject letter report, issued on January 31, 2008 to Wolf Creek Nuclear Operating Company for Wolf Creek, and subsequently again to Southern Nuclear Company for Vogtle Units 1 and 2 and to EXELON for Braidwood Unit 2 and Byron Unit 2, contains typographical errors in Section 5 of the report . In Section 5, all reference to Figures, Tables and References refer to section 6 which was removed from the report; all of the references to Figures, Tables and References should refer to section 5 of the report .

HOL CDC Author: H.O . Lagally Verified : C.D . Cassino Chemistry, Diagnostics Chemistry, Diagnostics and Materials Engineering and Materials Engineering Electronically approved records are authenticated in EDHS.

Attachment 7 Application for Steam Generator Tube Alternate Repair Criteria Technical Specification Amendment Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. 50-456 and 50-457 Exelon Generation Company, LLC (EGC) Commitments Related to the Implementation of the Steam Generator Tube Interim Alternate Repair Criteria Technical Specification Amendment

Attachment 7 Exelon Generation Company, LLC (EGC) Commitment Related to the Implementation of the Steam Generator Tube Interim Alternate Repair Criteria Technical Specification Amendment COMMITMENT TYPE COMMITTED DATE COMMITMENT ONE-TIME PROGRAMMATIC OR "OUTAGE" ACTION ACTION Yes/No (Yes/N21-For Braidwood Station Unit 2, EGC commits that for the tube end weld Prior to Mode 4 entry eddy current examination, the weld will during restart from be evaluated on a best effort basis for Braidwood Station Unit Yes No crack-like indications . Crack-like 2 Spring 2008 indications in the tube end weld that refueling outage exceed the maximum acceptable weld (A2R13) flaw size of 294 degrees will be visually examined on a best effort basis with inspection systems capable of achieving a resolution similar to the Maximum Procedure Demonstration Lower Case Character Height as discussed in ASME Section XI .

For Braidwood Station Unit 2, EGC Prior to Mode 4 entry commits that an evaluation for tube end during restart from welds that require plugging will be Braidwood Station Unit completed under the EGC Corrective 2 Spring 2008 Yes No Action Program to provide reasonable refueling outage assurance that unacceptable welds are (A2R13) removed from service .