L-09-172, License Amendment Request No. 07-007, Alloy 800 Steam Generator Tube Sleeving, Supplement

License Amendment Request No.07-007, Alloy 800 Steam Generator Tube Sleeving, Supplement

ML091980026
Person / Time
Site:	Beaver Valley
Issue date:	07/14/2009
From:	Sena P FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-09-172, TAC MD9969
Download: ML091980026 (26)

Text

FENOC Beaver Valley Power Station P.O. Box 4 FirstEnergyNuclearOperating Company Shippingport,PA 15077 Peter P. Sena III 724-682-5234 Site Vice President Fax: 724-643-8069 July 14, 2009 L-09-172 10 CFR 50.90 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 License Amendment Request No.07-007, Alloy 800 Steam Generator Tube Sleevinq, Supplement (TAC No. MD9969)

By letter dated October 10, 2008, (Reference 1), FirstEnergy Nuclear Operating Company (FENOC) requested an amendment to the operating license for Beaver Valley Power Station (BVPS) Unit No. 2. The proposed license amendment would modify the Technical Specifications to allow an additional method of repair for steam generator tubes involving the use of Westinghouse leak limiting Alloy 800 sleeves. The proposed amendment would also clarify an existing reporting requirement concerning steam generator tube inspection. By letter dated May 19, 2009, (Reference 2), the Nuclear Regulatory Commission (NRC) staff requested additional information to complete its review of the license amendment request.

By letter dated June 16, 2009, (Reference 3), FENOC submitted the response to Reference 2 that contains a regulatory commitment to revise the proposed changes to the Technical Specification submitted in Reference 1. This submittal fulfils regulatory commitment number 2 made in Reference 3. The proposed changes to the Technical Specification contained in this submittal are enhancements to those provided by Reference 3. The proposed changes to the Technical Specifications submitted by this letter supersede those submitted in Reference 1.

The information provided by this submittal does not invalidate the significant hazards consideration submitted by Reference 1 or the one published in the Federal Register on February 17, 2009.

As stated in Reference 1, FENOC requests approval of the proposed license amendment on or before October 12, 2009, corresponding to the start of BVPS Unit No. 2 refueling outage 2R14. Implementation is planned to occur prior to achieving Mode 4 during startup from the 2R14 refueling outage.

AI&W

Beaver Valley Power Station, Unit No. 2 L-09-172 Page 2 There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager -

Fleet Licensing, at 330-761-6071.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July /I/ , 2009.

Sincerely, Peter P. Sena III

References:

1. FENOC Letter L-08-307, "License Amendment Request No.07-007, Alloy 800 Steam Generator Tube Sleeving," dated October 10, 2008 (ADAMS Accession No.

MIL082890823)

2. NRC Letter dated May 19, 2009, "Beaver Valley Power Station, Unit No. 2 -

Request for Additional Information RE: The Alloy 800 Steam Generator Tube Sleeving License Amendment (TAC NO. MD9969)," (ADAMS Accession No. ML091350257)

3. FENOC Letter L-09-132, "Response to Request for Additional Information for License Amendment Request No.07-007, Alloy 800 Steam Generator Tube Sleeving (TAC NO. MD9969)," dated June 16, 2009

Enclosure:

FENOC Evaluation of the Supplement Proposed Changes cc: NRC Region I Administrator NRC Senior Resident Inspector NRR Project Manager Director BRP/DEP Site Representative (BRP/DEP)

Beaver Valley Power Station, Unit No. 2 License Amendment Request No.07-007 Supplement FENOC Evaluation of the Supplement Proposed Change

Subject:

Supplement to the application for amendment of Beaver Valley Power Station Unit No. 2 Technical Specifications to allow steam generator tube repair using leak limiting Alloy 800 sleeves.

Table of Contents Section Title Page 1.0 SUM MA RY DESC RIPTIO N............................................................. 1 2.0 D ETA ILED D ESC R IPTIO N ............................................................. 1 3.0 TEC HNICA L EVA LUATIO N ............................................................ 2

4.0 REGULATORY EVALUATION

................................................... 3 4.1 Significant Hazards Consideration ................................................... 3 4.2 Applicable Regulatory Requirements/Criteria .............................. 6 4 .3 P re ce de nt .................................................................................. .. 6 4 .4 C onclusio ns ............................................................................... . .6

5.0 ENVIRONMENTAL CONSIDERATION

...................................... 6 6.0 R E FE R E NC E S ........................................................................... 7 Attachments Number Title 1 Proposed Technical Specification Changes 2 Retyped Technical Specification Pages i

Beaver Valley Power Station Unit No. 2 License Amendment Request No.07-007 Supplement Page 1 of 7 1.0

SUMMARY

DESCRIPTION FirstEnergy Nuclear Operating Company (FENOC) proposes to amend Operating License NPF-73 for Beaver Valley Power Station (BVPS) Unit No. 2. The proposed changes would revise the Technical Specifications to allow the installation of leak limiting Alloy 800 sleeves as an additional approved steam generator (SG) tube repair method and would clarify an existing reporting requirement concerning SG tube inspection.

2.0 DETAILED DESCRIPTION The proposed Technical Specification changes, which are submitted for Nuclear Regulatory Commission (NRC) review and approval, are provided in Attachment 1.

Retyped Technical Specification replacement pages are provided in Attachment 2. The retyped replacement pages are provided to show the Technical Specification pages after the proposed changes have been incorporated and are labeled as "Unofficial" because the license amendment has not been issued and other BVPS amendments are expected to be issued prior to the issuance of the amendment requested in this submittal.

The proposed changes to the Technical Specifications have been prepared electronically. Deletions are shown with a strike-through, and insertions are shown double-underlined. This presentation allows the reviewer to readily identify the information that has been deleted and added. No changes are proposed for the Technical Specification Bases because the affected Technical Specifications are administrative and do not have associated Bases. To meet format requirements the index and Technical Specifications will be revised and repaginated as necessary to reflect the changes being proposed in this license amendment request (LAR) supplement.

This supplement submits a modification to one of the changes proposed in Reference 1 and two new proposed changes to the Technical Specifications. The changes proposed in this supplement were provided to the NRC for information in Reference 2 and are described in the following paragraphs.

Specification 5.5.5.2.d is changed by adding a note, shown below, to address the limited inspection capability demonstrated in the parent tube behind the nickel band.

The requirement for methods of inspection 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 does not apply to the portion of the original tube wall adjacent to the nickel band (the lower half) of the lower joint for the repair process that is discussed in Specification 5.5.5.2.f.3. However, the method of inspection in this area shall be a rotating plus point (or equivalent) coil.

The SG tube repair criterion of Specification 5.5.5.2.c.3 is applicable to flaws in this area.

Beaver Valley Power Station Unit No. 2 License Amendment Request No.07-007 Supplement Page 2 of 7 Specification 5.5.5.2.d is also changed by adding a new Specification (5.5.5.2.d.6),

shown below, to address inspection of the parent tube prior to sleeve installation. The addition of Specification 5.5.5.2.d.6 also requires a change to the first paragraph of Specification 5.5.5.2.d to reference d.6.

For Alloy 800 sleeves: The parent tube, in the area where the sleeve-to-tube hard roll joint (lower joint) and the sleeve-to-tube hydraulic expansion joint (upper joint) will be established, shall be inspected prior to installation of the sleeve.

Sleeve installation may proceed only if the inspection finds these regions free from service induced indications.

Specification 5.5.5.2.f.3 reflects a proposed change submitted in Reference 1 that is being modified in the supplement. The modification consists of adding a sentence to address the service life of the Alloy 800 sleeves. The proposed modification is shown below.

All Alloy 800 sleeves shall be removed from service by the spring of 2017 Unit 2 refueling outage (2R19).

The changes proposed in this supplement, and those proposed in Reference 1, are shown in Attachment 1. As a result, the proposed changes shown in Attachment 1 of this submittal supersede those submitted in Reference 1.

3.0 TECHNICAL EVALUATION

Since the technical evaluation of the originally proposed changes is provided in Reference 1, and remains valid for this submittal, the technical evaluation of only the changes proposed in this supplement, namely changes to Technical Specifications 5.5.5.2.d and 5.5.5.2.f.3, is provided in this section.

Specification 5.5.5.2.d describes required SG program provisions for SG tube inspections. It requires that the number and portions of 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. Since the nickel band of an Alloy 800 sleeve may interfere with detection of flaws of any type that may be present in the original tube adjacent to the nickel band, a note is provided that excludes the affected original tube region from this provision, but maintains a requirement to inspect the area and plug the tube if any flaw is detected. Additional details are provided in Enclosure D of Reference 1.

The addition of Specification 5.5.5.2.d.6 is made to address inspection of the parent tube prior to sleeve installation. Since the nickel band on transition zone sleeves presents challenges to standard eddy current examination of the parent tube behind the nickel band, the parent tube in the area of the sleeve-to-tube hard roll joint (lower joint) would be inspected using qualified inspection techniques prior to installation of the sleeve. This requirement is also carried over to the sleeve-to-tube hydraulically

Beaver Valley Power Station Unit No. 2 License Amendment Request No.07-007 Supplement Page 3 of 7 expanded joint (upper joint) as a remedial action. This new Specification will ensure that sleeve installation will proceed only if the inspection finds these regions free from service induced indications. Additional details are provided in Enclosures A and B of Reference 1.

The proposed modification to Section 5.5.5.2.f.3 limits the service life of Alloy 800 sleeves. This limitation is imposed because of the limited data available for in-service Alloy 800 sleeves and because once the sleeving technique has been approved, it will be applicable to only the existing SGs at BVPS Unit No. 2.

4.0 REGULATORY EVALUATION

FirstEnergy Nuclear Operating Company (FENOC) proposes to modify the Beaver Valley Power Station (BVPS) Unit No. 2 Technical Specifications to allow an additional method of repair for steam generator (SG) tubes that involve the use of Westinghouse leak limiting Alloy 800 sleeves, and clarify an existing reporting requirement concerning SG tube inspections. The significant hazards consideration provided in this supplement is a repeat of the one submitted by FENOC Letter L-08-307, "License Amendment Request No.07-007, Alloy 800 Steam Generator Tube Sleeving," dated October 10, 2008, (ADAMS Accession No. ML082890823). Therefore the following significant hazards consideration is consistent with the significant hazards consideration published in the Federal Register on February 17, 2009.

4.1 Significant Hazards Consideration FirstEnergy Nuclear Operating Company (FENOC) has evaluated whether or not a significant hazards consideration is involved with the proposed amendments 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. The leak limiting Alloy 800 sleeves are designed using the applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code and, therefore, meet the design objectives of the original steam generator (SG) tubing. The applied stresses and fatigue usage for the sleeves are bounded by the limits established in the ASME Code.

Mechanical testing has shown that the structural strength of sleeves under normal, upset, emergency, and faulted conditions provides margin to the acceptance limits. These acceptance limits bound the most limiting (three times normal operating pressure differential) burst margin recommended by Nuclear Regulatory Commission (NRC) Regulatory Guide 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes." Burst testing of sleeve/tube assemblies has confirmed the analytical results and demonstrated that no unacceptable levels of primary to secondary leakage are expected during any plant condition.

Beaver Valley Power Station Unit No. 2 License Amendment Request No.07-007 Supplement Page 4 of 7 The leak limiting Alloy 800 sleeve depth-based structural limit is determined using NRC guidance and the pressure stress equation of ASME Code,Section III with additional margin added to account for the configuration of long axial cracks. Calculations show that a depth-based limit of 45% through-wall degradation is acceptable. However, the proposed amendment provides additional margin by requiring an Alloy 800 sleeved tube to be plugged on detection of any flaw in the sleeve or in the pressure boundary portion of the original tube wall in the sleeve to tube joint. Degradation of the original tube adjacent to the nickel band of an Alloy 800 sleeve, at any depth, would not prevent the sleeve from satisfying design requirements. Thus, flaw detection capabilities within the original tube adjacent to the sleeve nickel band are not necessary in order to evaluate continued operation of the sleeved tube.

Evaluation of repaired SG tube testing and analysis indicates no detrimental effects on the leak limiting Alloy 800 sleeve or sleeved tube assembly from reactor system flow, primary or secondary coolant chemistries, thermal conditions or transients, or pressure conditions as may be experienced at BVPS Unit No. 2. Corrosion testing and historical performance of sleeve/tube assemblies indicates no evidence of sleeve or tube corrosion considered detrimental under anticipated service conditions.

Implementation of the proposed change has no significant effect on either the configuration of the plant or the manner in which it is operated. The consequences of a hypothetical failure of the leak limiting Alloy 800 sleeve/tube assembly is bounded by the current SG tube rupture (SGTR) analysis described in the BVPS Unit No. 2 Updated Final Safety Analysis Report because the total number of plugged SG tubes (including equivalency associated with installed sleeves) is required to be consistent with accident analysis assumptions. A main steam line break or feedwater line break would not cause a SGTR since the sleeves are analyzed for a maximum accident differential pressure greater than that predicted in the BVPS Unit No. 2 safety analysis. The sleeve/tube assembly leakage during plant operation would be minimal and is well within the allowable Technical Specification leakage limits and accident analysis assumptions, neither of which would be changed to compensate for the proposed repair method.

Proposed changes to Technical Specification 5.6.6.2.4 only affect a reporting requirement and do not affect plant design, operation or maintenance. They are editorial in nature and are intended as clarifications that would reinforce the original intent of the reporting requirement.

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

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

Beaver Valley Power Station Unit No. 2 License Amendment Request No.07-007 Supplement Page 5 of 7 Response: No. The leak limiting Alloy 800 sleeves are designed using the applicable ASME Code as guidance, and therefore meet the objectives of the original SG tubing. As a result, the functions of the SG will not be significantly affected by the installation of the proposed sleeve. Therefore, the only credible failure mode for the sleeve and/or tube is to rupture, which has already been evaluated. The continued integrity of the installed sleeve/tube assembly is periodically verified as required by the Technical Specifications and a sleeved tube will be plugged on detection of a flaw in the sleeve or in the pressure boundary portion of the original tube wall in the sleeve to tube joint.

Proposed changes to Technical Specification 5.6.6.2.4 only affect a reporting requirement and do not affect plant design, operation or maintenance. They are editorial in nature and are intended as clarifications that would reinforce the original intent of the reporting requirement.

Implementation of the proposed change has no significant effect on either the configuration of the plant, or the manner in which it is operated.

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

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

Response: No. The repair of degraded SG tubes with leak limiting Alloy 800 sleeves restores the structural integrity of the degraded tube under normal operating and postulated accident conditions. The reduction in reactor coolant system flow due to the addition of Alloy 800 sleeves is not significant because the cumulative effect of all repaired (sleeved) and plugged tubes will continue to be greater than the allowed reactor coolant system flow limit established in the Technical Specification LCO 3.4.1. The design safety factors utilized for the sleeves are consistent with the safety factors in the ASME Boiler and Pressure Vessel Code used in the original SG design.

Tubes with sleeves would also be subject to the same safety factors as the original tubes which are described in the performance criteria for SG tube integrity in the existing Technical Specifications. These performance criteria are not being changed to compensate for the proposed repair method. The sleeve and portions of the installed sleeve/tube assembly that represent the reactor coolant pressure boundary will be monitored and a sleeved tube will be plugged on detection of a flaw in the sleeve or in the pressure boundary portion of the original tube wall in the leak limiting sleeve/tube assembly. Use of the previously identified design criteria and design verification testing ensures that the margin of safety is not significantly different from the original SG tubes.

Beaver Valley Power Station Unit No. 2 License Amendment Request No.07-007 Supplement Page 6 of 7 Proposed changes to Technical Specification 5.6.6.2.4 only affect a reporting requirement and do not affect plant design, operation or maintenance. They are editorial in nature and are intended as clarifications that would reinforce the original intent of the reporting requirement.

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

Based on the above, FENOC concludes that the proposed amendments present 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.

4.2 Applicable Regulatory Requirements/Criteria Section 4.2 submitted in Reference 1 remains valid for this supplement.

4.3 Precedent Section 4.3 submitted in Reference 1 remains valid for this supplement.

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

5.0 ENVIRONMENTAL CONSIDERATION

A 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, or 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 amendment meets the eligibility criterion for categorical exclusion set forth 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.

Beaver Valley Power Station Unit No. 2 License Amendment Request No.07-007 Supplement Page 7 of 7

6.0 REFERENCES

1. FENOC Letter L-08-307, "License Amendment Request No.07-007, Alloy 800 Steam Generator Tube Sleeving," dated October 10, 2008, (ADAMS Accession No. ML082890823)

2. FENOC Letter L-09-132, "Response to Request for Additional Information for License Amendment Request No.07-007, Alloy 800 Steam Generator Tube Sleeving (TAC NO. MD9969)," dated June 16, 2009

Attachment 1 Beaver Valley Power Station, Unit No. 2 License Amendment Request No.07-007 Supplement Proposed Technical Specification Changes The following is a list of the affected pages:

5.5-6*

55.57*

5.5-8 5.5-9*

5.5-10 #

5.5-11 #

5.6-5*

5.6-6 No change. Page included for context only.

1. Changed by the Supplement.

Programs and Manuals5.5 No change. Page included for context only.

5.5 Programs and Manuals 5.5.5.1 Unit 1 Steam Generator (SG) Program (continued)

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

3. 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 interval between refueling outages (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.

e. Provisions for monitoring operational primary to secondary LEAKAGE 5.5.5.2 Unit 2 Steam Generator Program

a. 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 or repair of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, or repaired to confirm that the performance criteria are being met.

b. Provisions for 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.

1. 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 Beaver Valley Units 1 and 2 5.5-6 Amendments 278 / 161

Programs and Manuals No change. Page included for context only. 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 Steam Generator (SG) Program (continued) retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and, except for flaws addressed through application of the alternate repair criteria discussed in Specification 5.5.5.2.c.4, 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.

When alternate repair criteria discussed in Specification 5.5.5.2.c.4 are applied to axially oriented outside diameter stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than lx10-2.

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Except during a SG tube rupture, leakage from all sources excluding the leakage attributed to the degradation described in Specification 5.5.5.2.c.4 is also not to exceed 1 gpm per SG.

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational Leakage."

c. Provisions for SG Tube Repair Criteria

1. Tubes found by inservice inspection to contain a flaw in a non-sleeved region with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in Specification 5.5.5.2.c.4 or 5.5.5.2.c.5.

Beaver Valley Units 1 and 2 5.5- 7 Amendments 278 / 161

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 Steam Generator (SG) Program (continued)

2. Tubes found by inservice inspection to contain a flaw in a sleeve (excluding the sleeve to tube joint) with a depth equal to or exceeding the following percentages of the nominal sleeve wall thickness shall be plugged:

ABB Combustion Engineering TIG welded sleeves 27%

Westinghouse laser welded sleeves 25%

Westinghouse leak limiting Alloy 800 sleeves Any flaw

3. Tubes with a flaw in a sleeve to tube joint shall be plugged.

4. Tube support plate voltage-based repair criteria may be applied as an alternative to the 40% depth based criteria of Specification 5.5.5.2.c.1.

Tube Support Plate Plugging Limit is used for the disposition of an Alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is described below:

a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts will be allowed to remain in service.

b) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts will be repaired or plugged, except as noted in 5.5.5.2.c.4.c below.

c) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit (calculated according to the methodology in Generic Letter 95-05 as supplemented) may remain in service if a rotating pancake coil or acceptable alternative inspection does not detect degradation.

d) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the upper voltage repair limit (calculated according to the methodology in Generic Letter 95-05 as supplemented) will be plugged or repaired.

Beaver Valley Units 1 and 2 5.5 - 8 Amendments 278 6-1-TBD

Programs and Manuals No change. Page included for context only. 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 Steam Generator (SG) Proqram (continued) e) If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits specified in 5.5.5.2.c.4.a through 5.5.5.2.c.4.d.

The mid-cycle repair limits are determined from the following equations:

VMURL - SL 1.0+NDE+Gr ( CL-At)`

CL

)(CL -At),

CL VMLRL = VMURL (VURL - VLRL) where:

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

VSL = structural limit voltage Gr = average growth rate per cycle length NDE = 95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20-percent has been approved by NRC). The NDE is the value provided by the NRC in GL 95-05 as supplemented.

Implementation of these mid-cycle repair limits should follow the same approach as in Specifications 5.5.5.2.c.4.a through 5.5.5.2.c.4.d.

Beaver Valley Units 1 and 2 5.5-9 Amendments 278 / 161

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 Steam Generator (SG) Program (continued)

5. The F* methodology, as described below, may be applied to the expanded portion of the tube in the hot-leg tubesheet region as an alternative to the 40% depth based criteria of Specification 5.5.5.2.c.1:

a) Tubes with no portion of a lower sleeve joint in the hot-leg tubesheet region shall be repaired or plugged upon detection of any flaw identified within 3.0 inches below the top of the tubesheet or within 2.2 inches below the bottom of roll transition, whichever elevation is lower. Flaws located below this elevation may remain in service regardless of size.

b) Tubes which have any portion of a sleeve joint in the hot-leg Additional change proposed tubesheet region shall be plugged upon detection of any flaw by the Supplement. identified within 3.0 inches below the lower end of the lower sleeve joint. Flaws located greater than 3.0 inches below the lower end of the lower sleeve joint may remain in service regardless of size.

d. Provisions for SG Tube Inspections

-NOTE-The requirement for methods of inspection with the obiective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be Dresent along the length of the tube does not appyl to the porion of the original tube wall adiacent to the nickel band (the lower half) of the lower boint for the repair process that is discussed in Specification 5.5.5.2.f.3. However, the method of inspection in this area shall be a rotating plus point (or equivalent) coil. The SG tube repair criterion of Specification 5.5.5.2.c,3 is applicable to flaws in this area.

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 tubes repaired by sleeving, the portion of the original tube wall between the the requirements of d.1, d.2, d.3, d.4, d.__5 and d.65 below, the inspecto scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment 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.

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

2. Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one interval between refueling outages (whichever is less) without being inspected.

Beaver Valley Units 1 and 2 5.5- 10 Amendments 278 / 1-6-1-TBD

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 Steam Generator (SG) Program (continued)

3. 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 interval between refueling outages (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.

4. Indications left in service as a result of application of the tube support plate voltage-based repair criteria (Specification 5.5.5.2.c.4) shall be inspected by bobbin coil probe during all future refueling outages.

Implementation of the steam generator tube-to-tube support plate repair criteria requires a 100-percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20-percent random sampling of tubes inspected over their full length.

5. When the F* methodology has been implemented, inspect 100% of the inservice tubes in the hot-leg tubesheet region with the objective of Additional change proposed detecting flaws that may satisfy the applicable tube repair criteria of by the Supplement.

Specification 5.5.5.2.c.5 every 24 effective full power months or one interval between refueling outages (whichever is less).

6. For Alloy 800 sleeves: The Darent tube, in the area where the sleeve-to-tube hard roll joint (lower ioint) and the sleeve-to-tube hydraulic expansion ioint (upper ioint) will be established, shall be inspected prior to installation of the sleeve. Sleeve installation may proceed only if the inspection finds these regions free from service induced indications.

e. Provisions for monitoring operational primary to secondary LEAKAGE

f. Provisions for SG Tube Repair Methods Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.

Modification proposed by 1. ABB Combustion Engineering TIG welded sleeves, CEN-629-P, the Supplement. Revision 02 and CEN-629-P Addendum 1.

2. Westinghouse laser welded sleeves, WCAP-1 3483, Revision 2.

3. Westinghouse leak-limiting Alloy 800 sleeves, WCAP-15919-P, Revision 2. All Alloy 800 sleeves shall be removed from service by the spring of 2017 Unit 2 refueling outage (2R19).

Beaver Valley Units 1 and 2 5.5 - 11 Amendments 278 / 4-64TBD

Reporting Requirements No change. Page included for context only. 5.6 5.6 Reporting Requirements 5.6.6 Steam Generator Tube Inspection Report (continued) 5.6.6.2 Unit 2 SG Tube Inspection Report

1. A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.2, Unit 2 Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

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

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

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

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

h. The effective plugging percentage for all plugging and tube repairs in each SG, and

i. Repair method utilized and the number of tubes repaired by each repair method.

2. A report shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.2, Unit 2 Steam Generator Program, when voltage-based alternate repair criteria have been applied. The report shall include information described in Section 6.b of Attachment 1 to Generic Letter 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking."

3. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the Commission prior to returning the steam generators to service (MODE 4) should any of the following conditions arise:

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

Beaver Valley Units 1 and 2 5.6-5 Amendments 278 / 161

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6.2 Unit 2 Steam Generator Tube Inspection Report (continued)

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

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

4. Report the folowing infoFrnation

. o the NR A reDort shall be submitted within 90 days after aGhieVi-the initial entry into MODE 4 following an outage in which the F* methodology was applied. The report shall include the following hot-leg tubesheet region inspection results associated with the application of F*:

a. Total number of indications, location of each indication, orientation of each indication, severity of each indication, and whether the indications initiated from the inside or outside surface.

b. The cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet.

c. The projected end-of-cycle accident-induced leakage from tubesheet indications.

Beaver Valley Units 1 and 2 5.6- 6 Amendments 278 / 4-64TBD I

Attachment 2 Beaver Valley Power Station, Unit No. 2 License Amendment Request No.07-007 Supplement Retyped Technical Specification Changes The following is a list of the affected pages:

5.5-8 5.5-10 5.5-11 5.5-12 5.6-6

Unofficial Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 Steam Generator (SG) Progqram (continued)

2. Tubes found by inservice inspection to contain a flaw in a sleeve (excluding the sleeve to tube joint) with a depth equal to or exceeding the following percentages of the nominal sleeve wall thickness shall be plugged:

ABB Combustion Engineering TIG welded sleeves 27%

Westinghousd laser welded sleeves 25%

Westinghouse leak limiting Alloy 800 sleeves Any flaw I

3. Tubes with a flaw in a sleeve to tube joint shall be plugged.

4. Tube support plate voltage-based repair criteria may be applied as an alternative to the 40% depth based criteria of Specification 5.5.5.2.c.1.

Tube Support Plate Plugging Limit is used for the disposition of an Alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is described below:

a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts will be allowed to remain in service.

b) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts will be repaired or plugged, except as noted in 5.5.5.2.c.4.c below.

c) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit (calculated according to the methodology in Generic Letter 95-05 as supplemented) may remain in service if a rotating pancake coil or acceptable alternative inspection does not detect degradation.

d) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the upper voltage repair limit (calculated according to the methodology in Generic Letter 95-05 as supplemented) will be plugged or repaired.

Beaver Valley Units 1 and 2 5.5-8 Amendments 278 /TBD

I Unofficial Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 Steam Generator (SG) Program (continued)

5. The F* methodology, as described below, may be applied to the expanded portion of the tube in the hot-leg tubesheet region as an alternative to the 40% depth based criteria of Specification 5.5.5.2.c.1:

a) Tubes with no portion of a lower sleeve joint in the hot-leg tubesheet region shall be repaired or plugged upon detection of any flaw identified within 3.0 inches below the top of the tubesheet or within 2.2 inches below the bottom of roll transition, whichever elevation is lower. Flaws located below this elevation may remain in service regardless of size.

b) Tubes which have any portion of a sleeve joint in the hot-leg tubesheet region shall be plugged upon detection of any flaw identified within 3.0 inches below the lower end of the lower sleeve joint. Flaws located greater than 3.0 inches below the lower end of the lower sleeve joint may remain in service regardless of size.

d. Provisions for SG Tube Inspections

-NOTE-The requirement for methods of inspection with the objective of d6tecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube does not apply to the portion of the original tube wall adjacent to the nickel band (the lower half) of the lower joint for the repair process that is discussed in Specification 5.5.5.2.f.3. However, the method of inspection in this area shall be a rotating plus point (or equivalent) coil. The SG tube repair criterion of Specification 5.5.5.2.c.3 is applicable to flaws in this area.

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 tubes repaired by sleeving, the portion of the original tube wall between the sleeve's joints is not an area requiring re-inspection. In addition to meeting the requirements of d.1, d.2, d.3, d.4, d.5 and d.6 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. A degradation assessment 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.

Beaver Valley Units 1 and 2 5.5 - 10 Amendments 278 / TBD

Unofficial - Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 Steam Generator (SG) Program (continued)

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

2. Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one interval between refueling outages (whichever is less) without being inspected.

3. 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 interval between refueling outages (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.

4. Indications left in service as a result of application of the tube support plate voltage-based repair criteria (Specification 5.5.5.2.c.4) shall be inspected by bobbin coil probe during all future refueling outages.

Implementation of the steam generator tube-to-tube support plate repair criteria requires a 100-percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20-percent random sampling of tubes inspected over their full length.

5. When the F* methodology has been implemented, inspect 100% of the inservice tubes in the hot-leg tubesheet region with the objective of detecting flaws that may satisfy the applicable tube repair criteria of Specification 5.5.5.2.c.5 every 24 effective full power months or one interval between refueling outages (whichever is less).

6. For Alloy 800 sleeves: The parent tube, in the area where the sleeve-to-tube hard roll joint (lower joint) and the sleeve-to-tube hydraulic expansion joint (upper joint) will be established, shall be inspected prior to installation of the sleeve. Sleeve installation may proceed only ifthe inspection finds these regions free from service induced indications.

e. Provisions for monitoring operational primary to secondary LEAKAGE Beaver Valley Units 1 and 2 5.5- 11 Amendments 278 / TBD

Unofficial Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 Steam Generator (SG) Program (continued)

f. Provisions for SG Tube Repair Methods Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.

1. ABB Combustion Engineering TIG welded sleeves, CEN-629-P, Revision 02 and CEN-629-P Addendum 1.

2. Westinghouse laser welded sleeves, WCAP-1 3483, Revision 2.

3. Westinghouse leak-limiting Alloy 800 sleeves, WCAP-15919-P, Revision 2. All Alloy 800 sleeves shall be removed from service by the spring of 2017 Unit 2 refueling outage (2R19).

5.5.6 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables,

b. Identification of the procedures used to measure the values of the critical variables,

c. Identification of process sampling points,

d. Procedures for the recording and management of data,

e. Procedures defining corrective actions for all off control point chemistry conditions, and

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

Beaver Valley Units 1 and 2 5.5- 12 Amendments 278 / TBID

I Unofficial Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6.2 Unit 2 Steam Generator Tube Inspection Report (continued)

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

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

4. A report shall be submitted within 90 days after the initial entry into MODE 4 following an outage in which the F* methodology was applied. The report shall include the following hot-leg tubesheet region inspection results associated with the application of F*:

a. Total number of indications, location of each indication, orientation of each indication, severity of each indication, and whether the indications initiated from the inside or outside surface.

b. The cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet.

c. The projected end-of-cycle accident-induced leakage from tubesheet indications.

Beaver Valley Units 1 and 2 5.6- 6 Amendments 278 / TBD