ML081080044

Proposed License Amendment Request, Interim Alternate Repair Criteria (Iarc) for Steam Generator (SG) Tube Repair

ML081080044
Person / Time
Site:	Surry
Issue date:	04/14/2008
From:	Gerald Bichof Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
08-0207
Download: ML081080044 (53)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 April 14, 2008 10 CFR 50.90 U. S. Nuclear Regulatory Commission Serial No. 08-0207 ATTN: Document Control Desk SPS-LIC/PAK R1 Washington, D. C. 20555 Docket No. 50-281 License No. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT 2 PROPOSED LICENSE AMENDMENT REQUEST INTERIM ALTERNATE REPAIR CRITERIA (IARC) FOR STEAM GENERATOR (SG) TUBE REPAIR Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion) hereby requests an amendment in the form of changes to the Technical Specifications (TS) to Facility Operating License Number DPR-37 for Surry Power Station Unit 2. This amendment proposes a one cycle revision to Surry Power Station (SPS) Technical Specifications (TS). Specifically, TS 6.4.Q, "Steam Generator (SG) Program," and TS 6.6.3, "Steam Generator Tube Inspection Report," will be revised to incorporate an interim alternate repair criterion into the provisions for SG tube repair for use during the Surry Unit 2 2008 spring refueling outage and the subsequent operating cycle.

This amendment request is based, in part, on a similar request submitted by Wolf Creek Nuclear Operating Corporation (WCNOC) dated February 8, 2008 (Serial No. ET 08-0009). As part of their review of the WCNOC submittal, the NRC staff issued a request for additional information (RAI), and WCNOC responded to questions 1 through 5 of the RAI in a letter dated March 21, 2008 (Serial No. ET 08-0016). The NRC RAI and WCNOC's associated response have been addressed in the Dominion license amendment request where appropriate. The NRC also provided RAIs to other licensees who have requested similar license amendments. Westinghouse Electric Company responded, to these RAIs, and their response is included as Enclosure 4. This information has also been considered in Dominion's request, where appropriate.

Serial No. 08-0207 Docket No. 50-281 Page 2 of 5 Enclosure I provides the discussion of the proposed change. Enclosures 2 and 3 provide the marked-up and typed versions of the proposed TS pages, respectively. Enclosure 4 contains supporting technical information for the license amendment request prepared by Westinghouse Electric Company LLC (Westinghouse) and contains information proprietary to Westinghouse.

Therefore, this information is supported by affidavits, signed by Westinghouse, the owner of the information. The affidavits set forth the bases 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 2.390 of the Commission's regulations. Accordingly, it is respectfully requested that 'the information, which is proprietary to Westinghouse, be withheld from public disclosure in accordance with 2.390 of the Commission's regulations. The affidavits are included in Westinghouse authorization letters LTR-CAW-2411 and LTR-CAW-2412, "Application for Withholding Proprietary Information from Public Disclosure," which also include Proprietary Information Notices and Copyright Notices. The Westinghouse authorization letters are provided in Enclosure 6.

Correspondence with respect to the copyright or proprietary aspects of the Westinghouse information noted above or the supporting Westinghouse affidavits should reference the applicable authorization letter 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. Redacted, non-proprietary versions of the Westinghouse supporting documentation are provided in Enclosure 5.

We have evaluated the proposed amendment and have determined that it does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for our determination is included in Enclosure 1. We have also determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be released offsite and no significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change. The proposed change has been reviewed and approved by the Facility Safety Review Committee.

Dominion requests NRC approval of the proposed license amendment by May 14, 2008 to support the spring Surry Unit 2 refueling outage, which is currently scheduled to start in late April 2008. Once approved, the amendment will be implemented prior to increasing RCS temperature above 200OF during startup of Surry Unit 2 from the refueling outage.

Serial No. 08-0207 Docket No. 50-281 Page 3 of 5 If you have any questions or require additional information, please contact Mr. Gary D. Miller (804) 273-2771.

Sincerely, Gerald T. Bischof Vice President - Nuclear Engineering COMMONWEALTH OF VIRGINIA )

)

COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Gerald T. Bischof, who is Vice President

- Nuclear Engineering, of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this -Iday ofA -A. , 2008.

My Commission Expires: 3/ ,,0/0 Notary Public VICKI L. HUIl.

ICommonwooltth Of VKVWM lo cwdw J 106 31.4 2

Serial No. 08-0207 Docket No. 50-281 Page 4 of 5 Commitments made in this letter: None.

Enclosures:

1. Discussion of Change

2. Proposed Technical Specifications Pages (Marked-up)

3. Proposed Technical Specifications Pages (Typed)

4. Westinghouse Electric Company LLC Letters (Proprietary):

LTR-CDME-08-11 P-Attachment, "Interim Alternate Repair Criteria (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone," dated January 31, 2008.

LTR-CDME-08-43 P-Attachment, "Response to NRC Request for Additional Information Relating to LTR-CDME-08-11 P-Attachment," dated March 18, 2008.

5. Westinghouse Electric Company LLC Letters (Non-Proprietary):

Westinghouse Electric Company LLC, LTR-CDME-08-11 NP- Attachment "Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone," dated January 31, 2008.

Westinghouse Electric Company LLC, LTR-CDME-08-43 NP-Attachment "Response to NRC Request for Additional Information Relating to LTR-CDME-08-1 1-P- Attachment," dated March 18, 2008.

Westinghouse Electric Company LLC, LTR-CDME-08-25, "Errata for LTR-CDME-08-1 1; 'Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone,"' dated February 12, 2008.

Westinghouse Electric Company LLC, LTR-CDME-08-85, "Applicability of LTR-CDME-08-11 and LTR-CDME-08-043 to Surry Unit 1 and Unit 2,"

dated April 9, 2008.

6. Westinghouse Electric Company LLC Authorization Letters:

Westinghouse Electric Company LLC, LTR-CAW-08-241 1, "Application for Withholding Proprietary Information from Public Disclosure," dated April 9, 2008.

Westinghouse Electric Company LLC, LTR-CAW-08-2412, "Application for Withholding Proprietary Information from Public Disclosure," dated April 9, 2008.

Serial No. 08-0207 Docket No. 50-281 Page 5 of 5 cc: U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 NRC Senior Resident Inspector Surry Power Station State Health Commissioner Virginia Department of Health James Madison Building - 7 th Floor 109 Governor Street Room 730 Richmond, Virginia 23219 Mr. S. P. Lingam NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, Maryland 20852 Mr. R. A. Jervey NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, Maryland 20852

Serial No. 08-0207 Docket No. 50-281 ENCLOSURE 1 Discussion of Change Virginia Electric and Power Company (Dominion)

Surry Power Station Unit 2

Serial No. 08-0207 Docket No. 50-281 Enclosure I DISCUSSION OF CHANGE 1.0

SUMMARY

DESCRIPTION This amendment application proposes a one cycle revision to the Surry Power Station (SPS) Technical Specification (TS) 6.4.Q, "Steam Generator (SG) Program," and TS 6.6.A.3, "Steam Generator Tube Inspection Report," to incorporate an interim alternate repair criterion (IARC) into the provisions for SG tube repair criteria for use during the Surry Unit 2 2008 spring Refueling Outage 21 (RFO 21) and the subsequent operating cycle. This amendment application requests approval of an 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. This amendment application is required to preclude unnecessary SG tube plugging while still maintaining tube structural and leakage integrity.

2.0 DETAILED DESCRIPTION 2.1 Proposed Change The following specific changes to the Surry Units 1 and 2 TS are proposed:

• TS 6.4.Q - Steam Generator (SG) Program TS 6.4.Q.3 currently states:

"3. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged."

This criterion would be revised as follows, as noted in italic type:

"3. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following alternate tube repaircriteriashall be appliedas an alternative to the 40% depth-based criteria:

a. For Unit 2 Refueling Outage 21 and the subsequent operating cycle, tubes with flaws having a circumferential component less than or equal to 203 degrees found in the portion of the tube below 17 inches from the top of the tubesheet and above 1 inch from the bottom of the tubesheet do not require plugging. Tubes with flaws having a circumferential component greaterthan 1 of 15

Serial No. 08-0207 Docket No. 50-281 Enclosure 1 203 degrees found in the portion of the tube below 17 inches from the top of the tubesheet and above 1 inch from the bottom of the tubesheet shall be removed from service.

Tubes with service-inducedflaws located within the region from the top of the tubesheet to 17 inches below the top of the tubesheet shall be removed from service. Tubes with service-induced axial cracks found in the portion of the tube below 17 inches from the top of the tubesheet do not require plugging.

When more than one flaw with circumferential components is found in the portion of the tube below 17 inches from the top of the tubesheet and above I inch from the bottom of the tubesheet with the total of the circumferential components greater than 203 degrees and an axial separation distance of less than 1 inch, then the tube shall be removed from service. When the circumferentialcomponents of each of the flaws are added, it is acceptable to count the overlapped portions only once in the total of circumferential components.

When one or more flaws with circumferential components are found in the portion of the tube within 1 inch from the bottom of the tubesheet, and the total of the circumferential components found in the tube exceeds 94 degrees, then the tube shall be removed from service. When one or more flaws with circumferential components are found in the portion of the tube within 1 inch from the bottom of the tubesheet and within 1 inch axial separation distance of a flaw above 1 inch from the bottom of the tubesheet, and the total of the circumferential components found in the tube exceeds 94 degrees, then the tube shall be removed from service. When the circumferentialcomponents of each of the flaws are added, it is acceptable to count the overlappedportions only once in the total of circumferentialcomponents.

TS 6.6.A.3 - Steam Generator Tube Inspection Report TS 6.6.A.3 currently states:

"A report shall be submitted within 180 days after Tavg exceeds 200OF following completion of an inspection performed in accordance with the Specification 6.4.Q, 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, 2 of 15

Serial No. 08-0207 Docket No. 50-281 Enclosure 1

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

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

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

h. The effective plugging percentage for all plugging in each SG."

TS 6.6.A.3 would be revised to add the following three additional. reporting criteria:

J. Following completion of a Unit 2 inspection performed in Refueling Outage 21 (and any inspections performed in the subsequent operatingcycle), the number of indicationsand location, size, orientation,whether initiatedon primary or secondary side for each service-inducedflaw within the thickness of the tubesheet, and the total of the circumferentialcomponents and any circumferentialoverlap below 17 inches from the top of the tubesheet as determined in accordance with TS 6.4. Q.3.a,

j. Following completion of a Unit 2 inspection performed in Refueling Outage 21 (andany inspections performed in the subsequent operating cycle), the primary to secondary LEAKAGE rate observed in each steam generator(if it is not practicalto assign leakage to an individualSG, 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

k. Following completion of a Unit 2 inspection performed in Refueling Outage 21 (andany inspections performed in the subsequent operating cycle), the calculatedaccident leakage rate from the portion of the tube 17 inches below the top of the tubesheet for the most limiting accident in the most limiting steam generator.

2.2 Background TS 6.4.Q requires that a SG tube program be established and implemented to ensure that SG tube integrity is maintained. SG tube integrity is maintained by meeting specified performance criteria (in TS 6.4.Q) for structural and leakage integrity, consistent with the plant design and licensing bases. TS 6.4.Q requires a condition monitoring assessment be performed during each outage during which the SG tubes are inspected to confirm that the performance criteria are being met. TS 6.4.Q also includes provisions regarding the scope, frequency, and methods of SG tube inspections. Of relevance to the amendment application, these provisions require that the number and portions of tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type that may be present along the length of a tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-3 of 15

Serial No. 08-0207 Docket No. 50-281 Enclosure 1 tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria.

The applicable tube repair criteria, specified in TS 6.4.Q.3, are that tubes found by an inservice inspection to contain flaws with a depth equal to or exceeding 40 percent of the nominal tube wall thickness shall be plugged.

Reference 2 provides the technical justification for an 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 TTS is less than a value sufficient to permit the remaining circumferential ligament to transmit the limiting axial loads [the greater of 3 times the normal operating (NOP) loads or 1.4 times the steam line break (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 TTS.

The limiting circumferential ligament has been defined by calculation. 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 evaluation 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. The need to calculate leakage from individual cracks is avoided by the calculation of the ratio of accident induced leakage to normal operating leakage.

3.0 TECHNICAL EVALUATION

An evaluation has been performed in Reference 2 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:

1. Axial cracks in tubes below a distance of 17 inches from the TTS can remain in service in the Surry 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 (18-month SG tubing eddy current inspection interval).

After adjusting for the SG tube crack growth as documented in Reference 4, the allowable crack sizes in the tube (2030) 17 to 20 inches below TTS and (940) in the bottom 1 inch are bounding values and they apply for Model 51 F steam generators. The 4 of 15

Serial No. 08-0207 Docket No. 50-281 Enclosure 1 1.0 inch axial separation criterion discussed herein for multiple circumferential cracks likewise applies.

A bounding analysis approach is utilized for both the minimum ligament calculation and the 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; 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.

References and Tables in Section 5 of Reference 2 refer to the wrong section (e.g.,

Reference 6-1 should be 5-1). Westinghouse issued an errata letter to correct the discrepancies, and a copy of the letter is provided in Enclosure 5.

Questions Relating to IARC for Steam Generator Tubes This amendment request is based, in part, on a request for a similar TS amendment submitted by Wolf Creek Nuclear Operating Corporation (WCNOC) in Reference 1. The NRC provided a request for additional information (RAI) to WCNOC. In Reference 3 WCNOC responded to RAI questions 1 through 5. These RAI questions have been similarly addressed within this submittal for Surry Unit 2 where appropriate. The NRC also provided RAIs to other licensees who requested similar TS amendments. The responses to these additional questions (6 through 17) were provided in Reference 4 and have been incorporated into this request, where appropriate.

Discussion of Performance Criteria The following performance criteria of NEI 97-06, Rev. 2 (Reference 6), which are included in Surry's TS, are the basis for these analyses:

The structural integrity performance criterion is:

All in-service steam generatortubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 againstburst under normal steady state full power operationprimary to-secondary pressure differential and a safety factor of 1.4 againstburst applied to the design basis accident primary-to-secondarypressure 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 5 of 15

Serial No. 08-0207 Docket No. 50-281 Enclosure 1 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 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 SGs and leakage rate for an individual steam generator.Leakage is not to exceed I gpm per Steam Generator,except for specific types of degradationat specific locations when implementing alternate repair criteria as documented in the Steam GeneratorProgram technicalspecifications.

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 (e.g. 10 CFR 50.67).

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 Reference 2, the bounding remaining structural ligament which meets the NEI 97-06, Rev. 2, Performance Criterion described above and required for the tube to transmit the operational loads is 115 degrees arc. 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 Examination (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 6 of 15

Serial No. 08-0207 Docket No. 50-281 Enclosure 1 (Reference 7) describes the qualified technique used to detect circumferential primary water stress corrosion cracking (PWSCC) in the expansion transitions and in the tubesheet expansion zone (TEZ). 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 8). 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, 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 probes. Each probe has a "field of view," that is, a window of finite dimension in which it detects flaws. The field of view forý the + Point probe typically varies between 0.1 inch to 0.2 inch depending on the specific characteristics of the probe. Therefore, 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.

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Serial No. 08-0207 Docket No. 50-281 Enclosure 1 Consideration of Crack Growth The growth of cracks due to PWSCC in this submittal request is dictated by four default growth rates from Reference 2. The distribution of growth rates is assumed to be lognormal. Typical values and conservative values are given, although it is recommended in Reference 9 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 Reference'2 have mean values and 95% upper bound values. For this analysis, the typical 95% upper bound growth rate is used. The circumferential growth rates are expressed as inches per effective full power year (EFPY).

Table 1.0 Calculation of Required Minimum Ligament for 18 Months Operating Period Bounding EFPY Growth Growth Growth for Minimum Critical Structural (1) (In./EFPY) (Deg./EFPY) Operating Structural Ligament Ligament (2) (3) Period Ligament (degrees)

(degrees) (degrees) 18 1.5 0.12 20.65 31 115 146 Calendar Month (CM)

Operation

1) It is conservatively assumed that 1 EFPY= 1 Calendar Year

2) 95% upper value of typical growth rates from Reference (6)

3) Based on smallest (Model F) mean tubesheet bore dimension The residual structural ligament must be adjusted for growth during the anticipated operating period between the current and the next planned inspection. For the Surry SGs, referring to Table 1.0 above, the maximum allowable throughwall circumferential crack size in a SG tube is 2140 (=3600 - 1460) for one cycle of operation (18-month SG tubing eddy current inspection interval).

Note that the maximum allowable throughwall circumferential crack size in a SG tube was reduced to 2030 in the response to RAI question 11 in Reference 4. No additional uncertainty factor is necessary to be applied to the as-measured circumferential extent of the cracks.

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Serial No. 08-0207 Docket No. 50-281 Enclosure 1 Primary-to-Secondary Leakaqe Discussion Using the D'Arcy formula for flow through a porous medium, 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 if the observed leakage during normal operation is within its limits for the bounding plant is discussed in Reference 2. 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 the tubesheet; however, first principles dictate that the tubesheet bore and crevice must decrease. Therefore, the leakage analysis provided is conservative.

For the underlining assumptions of the IARC - no contact pressure between the tube and the tubesheet in the hydraulic expansion region - the discussion above shows that significant margins exist over the length of crevice required in a 17 inch span below the TTS. However, a conservative factor of 2.5 will be applied to that part of the observed normal operating leakage that cannot be associated with the degradation mechanisms outside the tubesheet expansion region to calculate the accident induced leakage from the tubesheet region (Reference 4).

For integrity assessments, the ratio of 2.5 will be used in completion of both the condition monitoring (CM) and operational assessment (OA) upon implementation of the IARC. For example, for the CM assessment, the component of leakage from the lower 4 inches for the most limiting steam generator during the prior cycle of operation will be multiplied by a factor of 2.5 and added to the total leakage from any other source and compared to the allowable accident analysis leakage assumption. For the OA, the difference in leakage from the allowable limit during the limiting design basis accident minus the leakage from the other sources will be divided by 2.5 and compared to the observed leakage. An administrative limit will be established to not exceed the calculated value.

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Serial No. 08-0207 Docket No. 50-281 Enclosure 1 Reporting Requirements Dominion is proposing the following reporting requirements:

" The proposed reporting requirements are only required for the applicable period of the IARC.

" Reference 2 determined the calculated accident leakage rate from the most limiting accident in the most limiting SG to be greater than 2 times the maximum operational primary to secondary leakage rate. Therefore, the reporting requirements do not include a requirement to describe how the calculated accident leakage rate from the most limiting accident was determined if the leakage rate is less than 2 times the maximum operational primary to secondary leakage rate.

Inspection and Repair of Tube The tube below the IARC depth will be examined with a qualified technique, e.g., +Point probe. Axial flaws have no impact on the structural integrity of the tube in this region and may be left in service. Circumferential indications that exceed the maximum acceptable tube flaw size of 203 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 bulge/overexpansion (BLG/OXP), the expansion may be limited to the non-inspected BLG/OXPs.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory RequirementslCriteria Steam Generator (SG) tube inspection and repair limits are specified in Section 6.4.Q, "Steam Generator (SG) Program" of the SPS 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 through wall. During the initial plant licensing of Surry Power Station Unit 2, it was demonstrated that the design of the reactor coolant pressure boundary met the regulatory 'requirements in place at that time. The General Design Criteria (GDC) included in Appendix A to 10 CFR Part 50 did not become effective until May 21, 1971. The Construction Permits for Surry Units 1 and 2 were issued prior to May 21, 1971; consequently, these units were not subject to GDC requirements.

(Reference SECY-92-223 dated September 18, 1992.) However, the following information demonstrates compliance with 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 10 of 15

Serial No. 08-0207 Docket No. 50-281 Enclosure 1 permit "periodic inspection and testing . . . to assess . . . structural and leak tight integrity" (GDC 32). Structural integrity refers to maintaining adequate margins against burst, and collapse of the steam generator tubing. Leakage integrity refers to limiting primary to secondary leakage during all plant conditions to within acceptable limits.

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. The reactor coolant pressure boundary is designed, fabricated and constructed so as to have an exceedingly low probability of gross rupture or significant uncontrolled leakage throughout its design lifetime. Reactor coolant pressure boundary components have provisions for the inspection testing and surveillance of critical areas by appropriate means to assess the structural and leaktight integrity of the boundary components during their service lifetime. Structural integrity refers to maintaining adequate margins against burst, and collapse of the steam generator tubing. Leakage integrity refers to limiting primary to secondary leakage during all plant conditions to within acceptable limits.

4.2 No Significant Hazards Consideration Dominion has evaluated whether 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:

(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) event evaluation and the postulated steam line break (SLB), and locked rotor 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 F steam generators has shown that axial loading of the tubes is negligible during an SSE.

At normal operating pressures, leakage from PWSCC below 17 inches from the UTS 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 203 degrees below 17 inches from the TTS for the subsequent operating cycle. Tube rupture is precluded for cracks in the hydraulic expansion 11 of 15

Serial No. 08-0207 Docket No. 50-281 Enclosure 1 region due to the constraint provided by the tubesheet. The potential for tube pullout is mitigated by limiting the allowable crack size to 203 degrees subsequent operating cycle. These allowable crack sizes take into account eddy current uncertainty and crack growth rate. It has been shown that a circumferential crack with an azimuthal extent of 203 degrees for the 18-month SG tubing eddy current inspection interval meet the performance criteria of NEI 97-06, Rev. 2, "Steam Generator Program Guidelines" and Draft Regulatory Guide (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes." 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 (including locked rotor) 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 150 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 470 gpd. This value is within the accident analysis assumptions for the limiting design basis accident for Surry, which is the postulated SLB event.

Based on the above, the performance criteria of NEI-97-06, Rev. 2 and Draft Regulatory Guide (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 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.

12 ofi5

Serial No. 08-0207 Docket No. 50-281 Enclosure 1 (3) Does the proposed chanqe involve a sigqnificant reduction in a mar-gin 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 limited tubesheet inspection depth 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 GDC 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-tubesheet weld, Reference 6 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 limited tubesheet inspection depth criteria will not result in unacceptable primary-to-secondary leakage during all plant conditions.

Based on the above, it is concluded that the proposed changes do not result in any reduction of margin with respect to plant safety as defined in the Updated Final Safety Analysis Report or bases of the plant Technical Specifications.

Therefore, Dominion 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.

4.3 Precedents Wolf Creek Nuclear Generating Station and Vogtle Electric Generating Plant, Units 1 and 2, were granted similar TS changes on April 4 and April 9,,2008, respectively.

These changes modified the inspection requirements for portions of the SG tubes greater than 17 inches below the top of the tubesheet.

4.4 Conclusion Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common 13 of 15

Serial No. 08-0207 Docket No. 50-281 Enclosure 1 defense and security or to the health and safety of the public with the implementation of the interim alternate repair criterion discussed above.

5.0 ENVIRONMENTAL CONSIDERATION

Dominion 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.

14 of 15

Serial No. 08-0207 Docket No. 50-281 Enclosure 1

6.0 REFERENCES

1. Letter from T. J. Garrett of Wolf Creek Nuclear Operating Corporation to USNRC dated February 8, 2008 (Serial No. ET 08-0009), "Docket No. 50-482: Revision to Technical Specification (TS) 5.5.9, 'Steam Generator (SG) Program' for Interim Alternate Repair Criteria."

2. Westinghouse Electric Company LLC Letter, LTR-CDME-08-1 1, "Interim Alternate Repair Criteria (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone," dated January 31, 2008.

3. Letter from T. J. Garrett of Wolf Creek Nuclear Operating Corporation to USNRC dated March 21, 2008 (Serial No. ET 08-0016), "Docket No. 50-482: Response to Request for Additional Information Related to License Amendment Request for an Interim Alternate Repair Criterion to Technical Specification 5.5.9, 'Steam Generator (SG) Program."'

4. Westinghouse Electric Company LLC Letter, LTR-CDME-08-43 P-Attachment "Response to NRC Request for Additional Information Relating to LTR-CDME 011 P-Attachment," dated March 18, 2008.

5. TSTF-449, Rev. 4,"Steam Generator Tube Integrity," Technical Specifications Task Force Standard Technical Specification Change Traveler, dated April 14, 2005.

6. NEI 97-06, Rev. 2, "Steam Generator Program Guidelines," May 2005.

7. ETSS #20510.1; Technique for Detection of Circumferential PWSCC at Expansion Transitions.

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

9. EPRI 1012987; "Steam Generator Integrity Assessment Guidelines," July 2006.

15 of 15

Serial No. 08-0207 Docket No. 50-281 ENCLOSURE 2 Proposed Technical Specifications Pages (Marked-up)

Virginia Electric and Power Company (Dominion)

Surry Power Station Unit 2

UO< kV eo 'fk3its }ae. TS 6.4-11 Q. Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

1. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

2. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

a. Structural integrity performance criterion: 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.

b. 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 1 gpm for all SG.

Amendment Nos. 251 and 250

TS 6.4-12

c. The operational LEAKAGE performance criterion is specified in TS 3.1.C and 4.13, "RCS Operational LEAKAGE."

3. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

4. 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. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of 4.a, 4.b, and 4.c 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.

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

b. Inspect 100% of the 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.

c. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

5. Provisions for monitoring operational primary to secondary LEAKAGE.

Amendment Nos. 4

INSERT A - Insert as new paragraph in TS 6.4.Q.3 on page TS 6.4-12:

The following alternate tube repair criteria shall be applied as an alternative to the 40% depth-based criteria:

a. For Unit 2 Refueling Outage 21 and the subsequent operating cycle, tubes with flaws having a circumferential component less than or equal to 203 degrees found in the portion of the tube below 17 inches from the top of the tubesheet and 1 inch from the bottom of the tubesheet do not require plugging. Tubes with flaws having a circumferential component greater than 203 degrees found in the portion of the tube below 17 inches from the top of the tubesheet and above 1 inch from the bottom of the tubesheet shall be removed from service.

Tubes with service-induced flaws located within the region from the top of the tubesheet to 17 inches below the top of the tubesheet shall be removed from service. Tubes with service-induced axial cracks found in the portion of the tube below 17 inches from the top of the tubesheet do not require plugging.

When more than one flaw with circumferential components is found in the portion of the tube below 17 inches from the top of the tubesheet and above 1 inch from the bottom of the tubesheet with the total of the circumferential components greater than 203 degrees and an axial separation distance of less than 1 inch, then the tube shall be removed from service. When the circumferential components of each of the flaws are added, it is acceptable to count the overlapped portions only once in the total of circumferential components.

When one or more flaws with circumferential components are found in the portion of the tube within 1 inch from the bottom of the tubesheet, and the total of the circumferential components found in the tube exceeds 94 degrees, then the tube shall be removed from service. When one or more flaws with circumferential components are found in the portion of the tube within 1 inch from the bottom of the tubesheet and within 1 inch axial separation distance of a flaw above 1 inch from the bottom of the tubesheet, and the total of the circumferential components found in the tube exceeds 94 degrees, then the tube shall be removed from service. When the circumferential components of each of the flaws are added, it is acceptable to count the overlapped portions only once in the total of circumferential components.

TS 6.6-3

b. The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.1 .D.4. In addition, the information itemized in Specification 3.1.D.4 shall be included in this report.

3. Steam Generator Tube Inspection Report A report shall be submitted within 180 days after Tavg exceeds 200'F following completion of an inspection performed in accordance with the Specification 6.4.Q, 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 during the inspection outage for each active degradation mechanism,

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

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

h. The effective plugging percentage for all plugging in each SG.

... 5.

Amendment Nos. '_=-and

INSERT B - Insert as new items in TS 6.6.A.3 on page TS 6.6-3:

i. Following completion of a Unit 2 inspection performed in Refueling Outage 21 (and any inspections performed in the subsequent operating cycle), the number of indications and location, size, orientation, whether initiated on primary or secondary side for each service-induced flaw within the thickness of the tubesheet, and the total of the circumferential components and any circumferential overlap below 17 inches from the top of the tubesheet as determined in accordance with TS 6.4.Q.3.a,

j. Following completion of a Unit 2 inspection performed in Refueling Outage 21' (and any inspections performed in the subsequent operating cycle), the primary to secondary LEAKAGE rate observed in each steam generator (if it is not practical to assign leakage to an individual SG, 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

k. Following completion of a Unit 2 inspection performed in Refueling Outage 21 (and any inspections performed in the subsequent operating cycle), the calculated accident leakage rate from the portion of the tube below 17 inches below the top of the tubesheet for the most limiting accident in the most limiting steam generator.

Serial No. 08-0207 Docket No. 50-281 ENCLOSURE 3 Proposed Technical Specifications Pages (Typed)

Virginia Electric and Power Company (Dominion)

Surry Power Station Unit 2

TS 6.4-12

c. The operational LEAKAGE performance criterion is specified in TS 3.1.C and 4.13, "RCS Operational LEAKAGE."

3. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following alternate tube repair criteria shall be applied as an alternative to the 40% depth-based criteria:

a. For Unit 2 Refueling Outage 21 and the subsequent operating cycle, tubes with flaws having a circumferential component less than or equal to 203 degrees found in the portion of the tube below 17 inches from the top of the tubesheet and 1 inch from the bottom of the tubesheet do not require plugging. Tubes with flaws having a circumferential component greater than 203 degrees found in the portion of the tube below 17 inches from the top of the tubesheet and above 1 inch from the bottom of the tubesheet shall be removed from service.

Tubes with service-induced flaws located within the region from the top of the tubesheet to 17 inches below the top of the tubesheet shall be removed from service. Tubes with service-induced axial cracks found in the portion of the tube below 17 inches from the top of the tubesheet do not require plugging.

When more than one flaw with circumferential components is found in the portion of the tube below 17 inches from the top of the tubesheet and above 1 inch from the bottom of the tubesheet with the total of the circumferential components greater than 203 degrees and an axial separation distance of less than I inch, then the tube shall be removed from service. When the circumferential components of each of the flaws are added, it is acceptable to count the overlapped portions only once in the total of circumferential components.

When one or more flaws with circumferential components are found in the portion of the tube within 1 inch from the bottom of the tubesheet, and the total of the circumferential components found in the tube exceeds 94 degrees, then the tube shall be removed from service. When one or more flaws with circumferential components are found in the portion of the tube within 1 inch from the bottom of the tubesheet and within 1 inch axial separation distance of a flaw above 1 inch from the bottom of the tubesheet, and the total of the circumferential components found in the tube exceeds 94 degrees, then the Amendment Nos.

TS 6.4-13 tube shall be removed from service. When the circumferential components of each of the flaws are added, it is acceptable to count the overlapped portions only once in the total of circumferential components.

4. 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. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of 4.a, 4.b, and 4.c 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.

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

b. Inspect 100% of the 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.

c. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

5. Provisions for monitoring operational primary to secondary LEAKAGE.

Amendment Nos.

TS 6.6-3

b. The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.1.D.4. In addition, the information itemized in Specification 3.1 .D.4 shall be included in this report.

3. Steam Generator Tube Inspection Report A report shall be submitted within 180 days after Tavg exceeds 200'F following completion of an inspection performed in accordance with the Specification 6.4.Q, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. ý 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 during the inispection outage for each active degradation mechanism,

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

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

h. The effective plugging percentage for all plugging in each SG.

i. Following completion of a Unit 2 inspection performed in Refueling Outage 21 (and any inspections performed in the'subsequent operating cycle), the number of indications and location, size, orientation, whether initiated on primary or secondary side for each service-induced flaw within the thickness of the tubesheet, and the total of the circumferential components and any circumferential overlap below 17 inches from the top of the tubesheet as determined in accordance with TS 6.4.Q.3.a, Amendment Nos.

TS 6.6-3a

j. Following completion of a Unit 2 inspection performed in Refueling Outage 21 (and any inspections performed in the subsequent operating cycle), the primary to secondary LEAKAGE rate observed in each steam generator (if it is not practical to assign leakage to an individual SG, 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

k. Following completion of a Unit 2 inspection performed in Refueling Outage 21 (and any inspections performed in the subsequent operating cycle), the calculated accident leakage rate from the portion of the tube below 17 inches below the top of the tubesheet for the most limiting accident in the most limiting steam generator.

Amendment Nos.

Serial No. 08-0207 Docket No. 50-281 ENCLOSURE 5 Westinghouse Electric Company LLC Letters (Non-Proprietary):

" Westinghouse Electric Company LLC, LTR-CDME-08-11 NP-Attachment "Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone," dated January 31, 2008

" Westinghouse Electric Company LLC, LTR-CDME-08-43 NP-Attachment "Response to NRC Request for Additional Information Relating to LTR-CDME-08-11-P- Attachment," dated March 18, 2008

• Westinghouse Electric Company LLC, LTR-CDME-08-25, "Errata for LTR-CDME-08-1 1; 'Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone,"' dated February 12, 2008

• Westinghouse Electric Company LLC, LTR-CDME-08-85, "Applicability of LTR-CDME-08-11 and LTR-CDME-08-043 to Surry Unit 1 and Unit 2,"

dated April 9, 2008 Virginia Electric and Power Company (Dominion)

Surry Power Station Unit 2

Westinghouse Proprietary Class 2 O Westinghouse To: P. J. McDonough Date: February 12, 2008 D. Alexander E. Arnold cc: 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-08-11; "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 Electronicallyapprovedrecords are authenticatedin EDMS.

WESTINGHOUSE PROPRIETARY CLASS 2

• Westinghouse To: D. Rogosky Date: April 9, 2008 R. Grendys cc: G. W. Whiteman J. A. Gresham E. P. Morgan C. D. Cassino J. T. Kandra From: H.O Lagally Your ref:

Ext: (724) 722-5082 Our ref: LTR-CDME-08-85 Fax: (724) 722-5909

Subject:

Applicability of LTR-CDME-08-11 and LTR-CDME-08-43 to Surry Unit 1 and Unit 2

References:

This letter contains LTR-CDME-08-1 1, "Interim Alternate Repair Criterion (ARC) for Cracks in the Lower Region of the Tubesheet Expansion Zone" and LTR-CDME-08-43, "Response to NRC Request for Additional Information (RAI) Relating to LTR-CDME-08-1 1 P-Attachment." Each of the letters includes the proprietary and non-proprietary version of the respective document. Together, these documents provide the technical basis for justification of an Interim Alternate Repair Criterion for the lower 4 inches of the tubesheet expansion region.

As a product of a jointly-funded effort among a number of utilities for the development of the IARC, the technical justification was developed as a bounding case for the affected plants with hydraulically expanded Alloy 600TT tubing, including Surry Unit I and Unit 2.

Therefore, the technical justification contained in these documents applies directly to Surry Units 1 and 2.

The Affidavits of Withholding for LTR-CDME-08-11 P and LTR-CDME-08-43, respectively will be transmitted under separate cover.

Please contact the undersigned should you have any questions or concerns.

Authors: Verifier:

• H. 0. Lagally, Fellow Engineer

• J. T. Kandra Chemistry, Diagnostics and Chemistry, Diagnostics and Materials Engineering Materials Engineering

• Electronicallyapprovedrecords are authenticatedin the electronic document management system.

Serial No. 08-0207 Docket No. 50-281 ENCLOSURE 6 Westinghouse Electric Company LLC Authorization Letters:

• Westinghouse Electric Company LLC, LTR-CAW-08-2411, "Application for Withholding Proprietary Information from Public Disclosure," dated April 9, 2008.

" Westinghouse Electric Company LLC, LTR-CAW-08-2412, "Application for Withholding Proprietary Information from Public Disclosure," dated April 9, 2008.

Virginia Electric and Power Company (Dominion)

Surry Power Station Unit 2

OWe.tinghous-e Westinghouse Electric Company Nuclear Services P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (412) 374-4011 Washington, DC 20555-0001 e-mail: greshaja@westinghouse.com Ourref: CAW-08-2411 April 9, 2008 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

LTR-CDME-08-43 P-Attachment, "Response to NRC Request for Additional Information Relating to LTR-CDME-08-11 P-Attachment," dated March 3, 2008 (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-08-2411 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 Dominion VA.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-08-241 1, 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 yo, rs, vJ.A. Gresham, Manager Regulatory Compliance and Plant Licensing Enclosures cc: Jon Thompson (NRC O-7E1A)

CAW-08-2411 bce: J. A. Gresham (ECE 4-7A) IL R. Bastien, 1L (Nivelles, Belgium)

C. Brinkman, IL (Westinghouse Electric Co., 12300 Twinbrook Parkway, Suite 330, Rockville, MD 20852)

RCPL Administrative Aide (ECE 4-7A) IL (letter and affidavit only)

G. W. Whiteman, Waltz Mill H. 0. Lagally, Waltz Mill C. D. Cassino, Waltz Mill J. T. Kandra, Waltz Mill D. L. Rogosky, ECE

CAW-08-2411 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:

J.~A.Gresham, Manager Regulatory Compliance and Plant Licensing Sworn to and subscribed before me this 9h day of April 2008 Notary Public COMMoNWENT4 OF PNNSYLVANIA I ~Notanali %we I Shran L Mara, Not*mFPu*

Monroeville Boro. Afegheny County MyCommisson Expim Jan. 29,2011 Member, Pennsylvania Association of Notaries

2 CAW-08-2411 (1) I am Manager, Regulatory Compliance and Plant Licensing, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I 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) I 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.

(3) I 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.

(i) 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 3 ~CAW-08-241 1

(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.

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

(f) 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-08-2411 (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.

(f) 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-08-43 P-Attachment, "Response to NRC Request for Additional Information (RAI) Relating to LTR-CDME-08-11 P-Attachment," dated March 3, 2008 (Proprietary), for submittal to the Commission, being transmitted by Dominion VA Application for Withholding Proprietary Information from Public Disclosure to the Document Control Desk. The proprietary information as submitted for use by Westinghouse for Surry Units 1 and 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 Surry Units 1 and 2 steam generators.

5 CAW-08-2411 (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.

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information

.so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the originalwas identified as proprietary.

• WeSt.ng ouse Westinghouse Electric Company Nuclear Services.

P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA U.S. Nuclear Regulatory Commission Directtel: (412) 374-4643 Document Control Desk Direct fax: (412) 374-4011 Washington, DC 20555-0001 e-mail: greshaja@westinghouse.com Our ref: CAW-08-2412 April 9, 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 Zone," 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-08-2412 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 Dominion VA.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-08-2412, 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 yo 5 J*.A. Gresham, Manager Regulatory Compliance and Plant Licensing Enclosures cc: Jon Thompson (NRC O-7E1A)

CAW-08-2412 bee: J. A. Gresham (ECE 4-7A) IL R. Bastien, 1L (Nivelles, Belgium)

C. Brinkman, IL (Westinghouse Electric Co., 12300 Twinbrook Parkway, Suite 330, Rockville, MD 20852)

RCPL Administrative Aide (ECE 4-7A) IL (letter and affidavit only)

G. W. Whiteman, Waltz Mill H. 0. Lagally, Waltz Mill C. D. Cassino, Waltz Mill J. T. Kandra, Waltz Mill D. L. Rogosky, ECE

CAW-08-2412 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:

J. A. Gresham, Manager Regulatory Compliance and Plant Licensing Sworn to and subscribed before me this 9 th day of April, 2008 Notary Public COMMONWEALTH OF PENNSYLVANIA Notaft &

Shamn L Marde, Noty Pubic Pny Boro, egheai MOemble y Conoty MY Commission Bow Jmeai 29, 2011 Member, Pennsylvania Association of Notaries

2 2 ~CAW-08-24 12 (1) 1 am Manager, Regulatory Compliance and Plant Licensing, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I 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.

(3) 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 2.3 90 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.

(i) 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-08-2412 (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.

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

(f) 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-08-2412 (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.

(f) 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 inpublic 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 Dominion VA Application for Withholding Proprietary Information from Public Disclosure to the Document Control Desk. The proprietary information as submitted for use by Westinghouse for Surry Units 1 and 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 Surry Units 1 and 2 steam generators.

5 5CAW-08-2412 (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.

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Serial No. 08-0207 Docket No. 50-281 THIS ATTACHMENT CONTAINS CONFIDENTIAL MATERIAL TO BE WITHHELD FROM PUBLIC DISCLOSURE PURSUANT TO 10 CFR 2.390 PROPRIETARY ENCLOSURE 4 Westinghouse Electric Company LLC Letters (Proprietary):

• LTR-CDME-08-11 P-Attachment, "Interim Alternate Repair Criteria (ARC) for Cracks in. the Lower Region of the Tubesheet Expansion Zone" dated January 31, 2008

• LTR-CDME-08-43 P-Attachment, "Response to NRC Request. for Additional Information Relating to LTR-CDME-08-11 P-Attachment,"

dated March 18, 2008 Virginia Electric and Power Company (Dominion)

Surry Power Station Unit 2

,h1