L-15-026, License Amendment Request to Steam Generator Technical Specifications

License Amendment Request to Steam Generator Technical Specifications

ML15092A569
Person / Time
Site:	Beaver Valley
Issue date:	04/01/2015
From:	Emily Larson FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-15-026
Download: ML15092A569 (57)

Text

Beaver Valley Power Station P.O. Box 4 FirstEnergy Nuclear Operating Company Shippingport, PA 15077 724-682-5234 Eric A. Larson Site Vice President Fax: 724-643-8069 April l, 2015 L-15-026 10 CFR 50.90 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit Nos. 1 and 2 Docket No. 50-334, License No. DPR-66 Docket No. 50-412, License No. NPF-73 License Amendment Request to Revise Steam Generator Technical Specifications Pursuant to 10 CFR 50.90 , FirstEnergy Nuclear Operating Company (FENOC) hereby requests an amendment to the facility operating licenses for Beaver Valley Power Station, Unit No. 1 and Unit No. 2 (BVPS-2). The proposed amendment would revise the following Technical Specifications associated with steam generators:

• 3.4.20, "Steam Generator (SG) Tube Integrity"

• 5.5 .5, "Steam Generator (SG) Program"

• 5.6.6, "Steam Generator Tube Inspection Report" Revision of Technical Specification 5.5.5 .2.f, "Provisions for SG Tube Repair Methods,"

is requested to support the use of Westinghouse leak-limiting Alloy 800 sleeves in the BVPS-2 SG tubes for five fuel cycles of operation. This change is requested as BVPS-2 SG replacement has been deferred . Additionally, inspection period changes, administrative changes, editorial corrections, and clarifications are proposed and include changes that are consistent with the guidance provided in Technical Specification Task Force Traveler TSTF-51 0, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection" (Accession No. ML110610350).

The FENOC evaluation of the proposed amendment is enclosed . Approval of the proposed amendment is requested by Apri14 , 2016. The amendment shall be implemented within 60 days of approval.

Beaver Valley Power Station, Unit Nos. 1 and 2 L-15-026 Page2 There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager- Fleet Licensing, at (330) 315-6810.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April ___1:_, 2015.

Sincerely, z-/-

?.- G!l/~

4' Eric A. Larson

Enclosure:

Evaluation of the Proposed Amendment cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative

Evaluation of the Proposed Amendment

Subject:

Proposed Revision of Technical Specification (TS) 3.4.20, "Steam Generator (SG) Tube Integrity"; TS 5.5.5, "Steam Generator (SG)

Program"; and TS 5.6.6, "Steam Generator Tube Inspection Report" for the Beaver Valley Power Station, Unit Nos. 1 and 2 Table of Contents 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Significant Hazards Consideration 4.2 Applicable Regulatory Requirements I Criteria 4.3 Precedent 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

Attachments

1. Proposed Changes to Technical Specifications, Annotated Copy

2. Proposed Changes to Technical Specifications, Retyped Copy

3. Proposed Changes to Technical Specification Bases, Annotated Copy

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 2 of 15 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Renewed Facility Operating License Nos. DPR-66 and NPF-73 for Beaver Valley Power Station Unit Nos. 1 (BVPS-1) and 2 (BVPS-2), respectively. Revision of Technical Specification (TS) 5.5.5.2.f, "Provisions for SG Tube Repair Methods," is requested to support the use of Westinghouse leak-limiting Alloy 800 sleeves in the BVPS-2 steam generator (SG) tubes for five fuel cycles of operation. This change is requested as BVPS-2 SG replacement has been deferred from the original planned date of spring of 2017 refueling outage (2R19).

Additionally, inspection period changes, administrative changes, editorial corrections, and clarifications toTS 3.4.20, "Steam Generator (SG) Tube Integrity,"

TS 5.5.5, "Steam Generator (SG) Program," and TS 5.6.6, "Steam Generator Tube Inspection Report," are proposed. Proposed changes to TSs 3.4.20, 5.5.5, and 5.6.6 apply to both BVPS-1 and BVPS-2 and include changes that are consistent with the guidance provided in Technical Specification Task Force Traveler TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection" (Agencywide Documents Access and Management System [ADAMS] Accession No. ML110610350). The availability of this TS improvement was announced in the Federal Register on October 27, 2011 (76 FR 66763) and is necessary to address implementation issues with respect to TSTF-449, Revision 4. TSTF-449, Revision 4 was implemented as Amendment No.

276 at BVPS-1 and Amendment No. 158 at BVPS-2 (ADAMS Accession No. ML062260011 ). Minor editorial corrections to the TSs are proposed as well.

Affected pages of the current TSs, annotated to show the proposed changes, are provided in Attachment 1. Re-typed TS pages with the proposed changes incorporated are provided in Attachment 2. TS Bases pages annotated to show proposed changes are provided for information in Attachment 3.

2.0 DETAILED DESCRIPTION 2.1 Proposed Change to TS 5.5.5.2.f, "Provisions for SG Tube Repair Methods" TS 5.5.5.2.f.3 currently requires all Westinghouse leak-limiting Alloy 800 sleeves in BVPS-2 SGs to be removed from service by the spring of 2017 refueling outage (2R19). The proposed change would allow the installation and use of Westinghouse leak-limiting Alloy 800 sleeves beyond the spring of 2017 refueling outage (2R 19),

but would not allow sleeve in-service lifetime to exceed the originally-approved five fuel cycles of operation.

On September 30, 2009, BVPS-2 License Amendment No. 170 (ADAMS Accession No. ML092590189) approved the use of Westinghouse leak-limiting Alloy 800 sleeves for SG tube repair, just prior to the fall of 2009 Unit 2 refueling outage

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 3 of 15 (2R14). Amendment No. 170 included a requirement for all Alloy 800 sleeves to be removed from service by the spring of 2017 Unit 2 refueling outage (2R19), which would limit the Alloy 800 sleeve operation to approximately seven-and-a-half years (five fuel cycles of operation).

BVPS-2 did not install Alloy 800 sleeves in the fall of 2009 refueling outage (2R14) or in the subsequent spring of 2011 refueling outage (2R15). The first use of Alloy 800 sleeves at BVPS-2 was fall of 2012 refueling outage (2R16). This proposed license amendment retains the existing requirement that an Alloy 800 sleeve shall remain in service for no more than five fuel cycles of operation; however, the limitation on having all sleeves removed from operation by the spring of 2017 refueling outage (2R19) is deleted.

2.2 Proposed Changes to TS 3.4.20, "Steam Generator (SG) Tube Integrity"; TS 5.5.5, "Steam Generator (SG) Program"; and TS 5.6.6, "Steam Generator Tube Inspection Report" TS 5.5.5, "Steam Generator (SG) Program," provides requirements to establish and implement a program to ensure that SG tube integrity is maintained. TS 5.5.5 contains the Steam Generator Program for Unit 1 (TS 5.5.5.1) and the Steam Generator Program for Unit 2 (TS 5.5.5.2). The unit-specific programs are due to different SGs in use at each unit. The original BVPS-1 SGs have been replaced with Westinghouse model 54F SGs with Alloy 690 thermally-treated (690TT) tube material, while the original BVPS-2 SGs are Westinghouse modei51M with Alloy 600 mill-annealed (600MA) tube material. The programs are also different due to the provisions for SG tube repair methods in BVPS-2 Specification 5.5.5.2.f (BVPS-1 does not have an approved SG tube repair method). However, both programs include:

• Condition monitoring assessments (TSs 5.5.5.1.a I 5.5.5.2.a)

• Performance criteria for SG tube integrity (TSs 5.5.5.1.b 15.5.5.2.b)

• SG tube repair criteria (TSs 5.5.5.1.c I 5.5.5.2.c)

• SG tube inspections (TSs 5.5.5.1.d I 5.5.5.2.d)

• Monitoring operational primary to secondary LEAKAGE (TSs 5.5.5.1.e I 5.5.5.2.e)

Consistent with TSTF-51 0, Revision 2, an editorial correction is made to Paragraph 5.5.5.1.b.1 and 5.5.5.2.b.1. The closing parenthesis is misplaced and inappropriately includes anticipated transients in the description of normal operating conditions.

Consistent with TSTF-51 0, Revision 2, the title for Paragraph 5.5.5.1.c is revised to "Provisions for SG Tube Plugging Criteria" as BVPS-1 has no approved repair methods. The title for Paragraph 5.5.5.2.c is revised to "Provisions for SG Tube Plugging or Repair Criteria" as BVPS-2 may plug SG tubes or repair with approved

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 4 of 15 repair methods. References to tube repair are changed to "plugging or repair," or "plugged or repaired" throughout Specification 5.5.5.2. To be consistent with these changes, references to "tube repair criteria" are revised to "tube plugging or repair criteria" in the Steam Generator (SG) Tube Integrity Specification 3.4.20 and the associated Bases.

Consistent with TSTF-51 0, Revision 2, clarifications are made to Paragraph 5.5.5.1.d. The reference to "tube repair criteria" is changed to "tube plugging criteria" as BVPS-1 has no approved repair methods.

TS 5.5.5.1.d.1 is revised from "Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement" to "Inspect 100% of the tubes in each SG during the first refueling outage following SG installation" to be consistent with TSTF-510, Revision 2.

TS 5.5.5.1.d.2 is modified consistent with the TSTF-510, Revision 2 insertion for the 690TT SG tubes used in BVPS-1. The proposed change modifies the frequency of verification of SG tube integrity and SG tube sample selection to reduce implementation issues experienced with the current specification. The revised specification is consistent with the existing specification in that it continues to be based on SG tube material type, age, condition and cycle length; continues to address the time dependence of degradation; and prevents front end or back end loading of inspections. The maximum interval allowed between inspections remains the same as in the current TSs. In addition, the proposed change addresses an administrative inconsistency in TSTF-510, Paragraph d.2 of the Steam Generator Tube Inspection Program. "Tube repair criteria" should be "tube plugging criteria" for BVPS-1. This administrative error was acknowledged in an NRC letter dated June 17, 2013 (ADAMS Accession No. ML13120A541).

TS 5.5.5.1.d.3 is modified consistent with TSTF-51 0, Revision 2 to clarify the inspection requirements when crack indications are found in any SG tube and the intent of the parenthetical statement. The wording " ... shall not exceed 24 effective full power months or one interval between refueling outages" is changed to " ... shall not exceed 24 effective full power months or one refueling outage" to be consistent with Specification 5.5.9.d.3 of Standard Technical Specifications- Westinghouse Plants (NUREG-1431, Revision 4, ADAMS Accession No. ML12100A222).

Consistent with TSTF-51 0, Revision 2, a change to the note for paragraph 5.5.5.2.d is made to clarify that for flaws in this particular inspection area, the plugging criterion of Specification 5.*5.5.2.c.3 is applicable instead of the repair criterion of Specification 5.5.5.2.c.3.

TS 5.5.5.2.d.1 is revised from "Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement," to "Inspect 100% of the tubes in each

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 5 of 15 SG during the first refueling outage following SG installation," to be consistent with TSTF-510, Revision 2.

TS 5.5.5.2.d.2 is modified consistent with the TSTF-510, Revision 2 insertion for the 600MA SG tubes used in BVPS-2. The proposed change modifies the frequency of verification of SG tube integrity and SG tube sample selection to reduce implementation issues experienced with the current specification. The revised specification is consistent with the existing specification in that it continues to be based on SG tube material type, age, condition and cycle length; continues to address the time dependence of degradation; and prevents front end or back end loading of inspections. The maximum interval allowed between inspections remains the same as in the current TSs.

TS 5.5.5.2.d.3 is deleted as 5.5.5.2.d.2 already requires a SG inspection interval of 24 effective full power months or at least every refueling outage (whichever results in more frequent inspections). This is consistent with the note on page 13 of the NRC model safety evaluation for TSTF-510, Revision 2 (ADAMS Accession No. ML112101513). Subsequent 5.5.5.2.d paragraphs are renumbered, and the number of requirements updated in TS 5.5.5.2.d, "Provisions for SG Tube Inspections."

Consistent with TSTF-51 0, Revision 2, clarifications are made to Specification 5.6.6, "Steam Generator Tube Inspection Report" for both BVPS-1 and BVPS-2. The required effective plugging percentage in each SG from paragraph h of TS 5.6.6.1 and TS 5.6.6.2.1 is moved to paragraph f. The term "active" is deleted from paragraphs b and e of TS 5.6.6.1 and paragraphs b and e of TS 5.6.6.2.1 to be consistent with TS 5.5.5.

An editorial correction is made to TS 5.5.5.2.b.3 where LCO 3.4.13 "RCS Operational Leakage" is changed to "RCS Operational LEAKAGE." An editorial change is made by moving the word "and" to the end of TS 5.6.6.1.f from the end of TS 5.6.6.1.g, since the last item in the series (TS 5.6.6.1.h) was deleted and the requirements moved. A similar editorial change is made toTS 5.6.6.2.1.g.

An editorial change is made to the TS 5.5.5.1 and 5.5.5.2 titles to replace "Steam Generator" with "SG."

An editorial change toTS 5.6.6 "Steam Generator Tube Inspection Report" is made to insert the acronym (SG) in the title. Also, the Specification 5.5.5.1 and 5.5.5.2 titles included in Specifications 5.6.6.1, 5.6.6.2.1, and 5.6.6.2.2 are updated by changing "Steam Generator" to "SG." Quotation marks are also inserted in these three titles.

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 6 of 15

3.0 TECHNICAL EVALUATION

The SGs in pressurized water reactor designs remove heat from the reactor coolant system (RCS) and produce steam to operate the main generator and other balance-of-plant equipment. SG tubes constitute the heat transfer surface area between the primary (reactor coolant) and secondary (main steam) systems and, as such, are relied on to maintain the primary system's pressure and inventory. As an integral part of the reactor coolant pressure boundary (RCPB), the SG tubes isolate the radioactive fission products in the primary coolant from the secondary system in the SGs. Maintaining tube integrity ensures that the tubes are capable of performing their intended safety functions consistent with the plant licensing basis and applicable regulatory requirements.

The licensing basis for BVPS-1 and BVPS-2 includes the postulation of a SG tube rupture (SGTR) accident. In the event of a SGTR, primary coolant is released into the secondary side of the SG and subsequently can be released to the environment through main steam safety valves, atmospheric dump valves, or leak paths in the secondary system. A SGTR is a design basis accident for which analyses are summarized in section 14.2.4, "Steam Generator Tube Rupture" of the BVPS-1 Updated Final Safety Analysis Report (UFSAR), and section 15.6.3, "Steam Generator Tube Rupture (SGTR)" of the BVPS-2 UFSAR.

In order to ensure that the probability of a SGTR does not increase above that assumed in the accident analysis and that no other design basis accidents or transients result in tube failure, it is necessary to maintain SG tube integrity. For that purpose, TS 5.5.5, "Steam Generator (SG) Program" imposes requirements for monitoring, inspection, and maintenance to ensure SG tube integrity remains consistent with licensing basis assumptions related to SGTR and other design basis accidents and transients.

3.1 Proposed Changes to Technical Specification 5.5.5.2.f, "Provisions for SG Tube Repair Methods" The proposed changes would revise Technical Specification 5.5.5.2.f.3 to ensure that a Westinghouse leak-limiting Alloy 800 sleeve used for SG tube repair at BVPS-2 shall remain in service for no more than five fuel cycles of operation starting from the outage when the sleeve was installed.

FirstEnergy Nuclear Operating Company (FENOC) originally applied for the use of Westinghouse leak-limiting Alloy 800 sleeves to repair SG tubes at BVPS-2 by letter dated October 10, 2008 (ADAMS Accession No. ML082890823). FENOC provided additional information, in response to an NRC request for additional information, by letters dated June 16, 2009 and July 14, 2009 (ADAMS Accession Nos.

ML091690044 and ML091980026, respectively).

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 7 of 15 The technical information, bases, and letter attachments previously provided continue to remain applicable and valid for BVPS-2. The information is again credited herein to support this proposed change, except for the administrative calendar statements involving the Alloy 800 sleeve start-use date projection and the SG replacement date projection, as described below.

The NRC previously approved the use of Westinghouse leak-limiting Alloy 800 sleeves for use in BVPS-2 SG tube repair by Amendment No. 170. Concerns were noted in three areas of the Safety Evaluation for corrosion of the sleeve and sleeve/tube assembly (section 3.4.2), inspection of the parent tube behind the nickel band portion of the sleeve (section 3.5), and structural degradation of the sleeve (section 3.6). Besides testing and analyses, these three concerns were addressed by periodic inspection requirements and limiting sleeve service life:

• Section 3.4.2, "Corrosion Testing," of Amendment No. 170 states in part,

" ... at present, the NRC staff can only assume a limited life expectancy for leak-limiting Alloy 800 sleeves. Considering the uncertainties in sleeve life expectancy, sleeves are periodically inspected to ensure any flaws in the sleeve/tube assembly are detected and addressed."

• Section 3.5 "Sleeve Inspection," of Amendment No. 170 states in part,

"... the licensee has limited the amount of time that the sleeves will be in service by proposing a TS requirement to remove all Alloy 800 sleeves from service by the spring of 2017 BVPS-2 refueling outage (2R19). The limitation on the service life of the sleeve limits the amount of time that degradation of the sleeve joint could occur."

• Section 3.6 "Sleeve Structural Analysis," of Amendment No. 170 states in part, "The calculated amount of degradation that could be tolerated and still meet ASME limits was considered acceptable to the NRC since degradation of the sleeve is unlikely for the period of time the sleeve will be inservice, (less than 8 years), and the licensee will plug all flaws on detection."

The sleeve service life was discussed in the June 16, 2009 request for additional information response. The following is an excerpt from the response.

Replacement of the BVPS Unit No. 2 steam generators is currently scheduled for the spring of 2017 refueling outage (2R 19). Reference 1 requested approval of the Alloy 800 sleeve to support the fall of 2009 refueling outage (2R14). Therefore, the service life of the Alloy 800 sleeve should be expected not to exceed five operating cycles if installed in the fall of 2009. This sleeve design is not applicable to the replacement steam generators and would require a separate license amendment request to install the sleeve in the replacement steam generators.

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 8 of 15 A new section (5.5.5.2.f.3) will be added to the BVPS Technical Specifications to address the service life of the Alloy 800 sleeves.

The proposed wording for Technical Specification 5.5.5.2.f.3 is:

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

The June 16, 2009 request for additional information response indicated that:

1) Approval to use Alloy 800 sleeves was requested to support the BVPS-2 fall of 2009 refueling outage (2R14); and 2) replacement of the BVPS-2 steam generators was scheduled to occur in the spring of 2017 refueling outage (2R19). The first use of Alloy 800 sleeves did not occur in accordance with that timeline. No Alloy 800 sleeves were installed at BVPS-2 in either the fall of 2009 refueling outage (2R14) or the spring of 2011 refueling outage (2R15). The first use of Alloy 800 sleeves at BVPS-2 occurred in the fall of 2012 refueling outage (2R 16). Replacement of the BVPS-2 steam generators is currently scheduled to occur in the spring of 2020 refueling outage (2R21).

The proposed changes toTS 5.5.5.2.f.3 retain the current credited service life basis limitation that an Alloy 800 sleeve shall remain in service for no more than five fuel cycles of operation starting from the outage when the sleeve was installed (approximately seven-and-a-half years of service life).

BVPS-2 TS 5.5.5.2.d continues to require periodic SG tube inservice inspections.

Furthermore, additional operating experience, as discussed below, has occurred since Alloy 800 sleeves were first approved in Amendment No. 170 that continues to support Alloy 800 sleeve service life.

No degradation of the sleeves installed at BVPS-2 has been reported. There are 94 Alloy 800 sleeves currently in service at BVPS-2.

As of March 2015, more than 18,000 Alloy 800 sleeves have been installed both domestically and internationally. Of those sleeves still in the operating SGs in which they were originally installed, FENOC is not aware of any sleeve degradation or degradation of the parent tube in the joint regions being reported. The phenomenon of trapped fluid causing inward deformation of tube sleeves has not been observed to date in the leak-limiting Alloy 800 sleeve design.

In conclusion, the BVPS-2 TS Amendment No. 170 limitation that an Alloy 800 sleeve shall remain in service for no more than five fuel cycles of operation (approximately seven-an-a-half years) continues to be applicable. The proposed change is that the requirement to have all Alloy 800 sleeves at BVPS-2 removed from service by the spring of 2017 refueling outage (2R 19) is deleted.

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 9 of 15 3.2 Proposed Changes to Technical Specifications 3.4.20, "Steam Generator (SG)

Tube Integrity," 5.5.5, "Steam Generator (SG) Program," and 5.6.6, "Steam Generator Tube Inspection Report."

TSTF-510, Revision 2, provides revisions to the SG Program TSs that stipulate the SG tube inspections and inspection frequencies and have been determined to be appropriate for the BVPS-1 and BVPS-2 SGs. TSTF-51 0, Revision 2, includes separate sections for the 690TT SG tubing used at BVPS-1 and the 600MA SG tubing used at BVPS-2. The proposed changes toTS 5.5.5 directly reflect the wording provided in TSTF-510, Revision 2 for the appropriate SG tubing materials in use at BVPS-1 and BVPS-2.

Current TS 5.5.5 references SG tube repair criteria. TSTF-510, Revision 2, specifies the reference to SG tube repair options be deleted if no repair methods are approved. BVPS-1 does not have approved repair methods while BVPS-2 has approved repair methods. The revised specifications for BVPS-1 and BVPS-2 directly reflects the wording provided in TSFTF-51 0, Revision 2 for SG tube repair options.

Current TS 3.4.20 establishes the requirement to maintain SG tube integrity in MODEs 1, 2, 3, and 4 and establishes the requirements for addressing SG tubes that satisfy tube repair criteria through either tube plugging or repair in accordance with the Steam Generator Program. Consistent with TSTF-510, Revision 2 and the proposed changes to TS 5.5.5 described above, Limiting Condition for Operation (LCO) 3.4.20, and its associated ACTIONS and SURVEILLANCE REQUIREMENTS, would be revised to reference the tube plugging or repair criteria in accordance with the Steam Generator Program. The revised specification directly reflects the wording provided in TSTF-510, Revision 2.

The proposed changes toTS 5.6.6 directly reflect the changes specified in TSTF-51 0, Revision 2, to clarify the reporting requirements. These changes have no impact on plant operation or SG inspection requirements.

The editorial TS corrections for acronyms, capitalization, and added quotation marks do not change the intent of the requirements.

The proposed changes do not affect the design of the SGs, their method of operation, the operational leakage limit, the accident analyses or primary coolant chemistry controls. The proposed changes are an improvement to the existing SG inspection requirements and contains a number of editorial corrections, changes, and clarifications intended to improve internal consistency, consistency with the implementing industry documents, and usability without changing the intent of the requirements.

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 10 of 15

4.0 REGULATORY EVALUATION

FirstEnergy Nuclear Operating Company (FENOC) is requesting amendment of Renewed Facility Operating License Nos. DPR-66 and NPF-73 for Beaver Valley Power Station Unit Nos. 1 (BVPS-1) and 2 (BVPS-2), respectively. The amendment would revise Technical Specifications (TSs) 3.4.20, "Steam Generator (SG) Tube Integrity," 5.5.5, "Steam Generator (SG) Program," and 5.6.6, "Steam Generator Tube Inspection Report." A proposed change toTS 5.5.5.2, "Unit 2 Steam Generator Program" is requested to permit the use of Alloy 800 sleeves in the BVPS-2 SG tubes for five fuel cycles of operation. This change would allow the full five-cycle Alloy 800 sleeve service life as originally permitted in Amendment No. 170 and is requested due to deferred SG replacement.

Additionally, proposed changes are needed to address implementation issues associated with the inspection periods and make other administrative changes, editorial corrections, and clarifications. The majority of these changes are consistent with the guidance provided in Technical Specification Task Force Traveler No. 510 (TSTF-51 0), Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," that was approved by the Nuclear Regulatory Commission staff on October 27, 2011. The proposed changes are an improvement to the existing SG inspection requirements and contains a number of editorial corrections, changes, and clarifications intended to improve internal consistency, consistency with the implementing industry documents, and usability without changing the intent of the requirements.

4.1 Significant Hazards Consideration FENOC has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No The proposed changes to Technical Specification 5.5.5.2.f.3 replaces the date and outage when all Alloy 800 sleeves shall be removed from service with a limitation on the individual sleeve service life from the date of installation. The allowed maximum service life previously approved for Alloy 800 sleeves remains unchanged. Since the maximum service life of the Alloy 800 sleeves is unchanged, the probability of a failure due to degradation does not increase.

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 11 of 15 Implementation of the proposed changes toTS 5.5.5.2.f.3 have 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 are bound by the current steam generator tube rupture (SGTR) analysis described in the BVPS-2 Updated Final Safety Analysis Report (UFSAR) 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-2 accident 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 repair method.

The proposed changes to TSs 3.4.20, 5.5.5, and 5.6.6 are consistent with TSTF-510, editorial corrections, and clarifications. Changes that are consistent with TSTF-51 0 and other editorial corrections and clarifications do not change the physical plant or how it is operated; therefore they cannot affect the probability or consequence of a previously-evaluated accident. A proposed change modifies the frequency of verification of SG tube integrity and SG tube sample selection. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed changes will not cause the consequences of a SGTR to exceed those assumptions. Therefore, it is concluded that these changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No Proposed changes to Technical Specification 5.5.5.2.f.3 replaces the date and outage when all Alloy 800 sleeves shall be removed from service with a limitation on the individual sleeve service life from the date of installation. The allowed maximum service life previously approved for Alloy 800 sleeves remains unchanged.

Implementation of these proposed changes have no significant effect on either the configuration of the plant or the manner in which it is operated.

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 12 of 15 The leak-limiting Alloy 800 sleeves are designed using the applicable ASME Code as guidance and 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 or tube is to rupture, which has already been evaluated. No new failure modes, malfunctions, or accident initiators have been created. 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.

The proposed changes to TSs 3.4.20, 5.5.5, and 5.6.6 are changes*

consistent with TSTF-51 0, editorial corrections, and clarifications. These changes do not affect the operation of the SGs or the ability of the SGs to perform their design or safety functions; therefore they do not create new failure modes, malfunctions, or accident initiators.

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

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

Response: No The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.

Proposed changes to Technical Specification 5.5.5.2.f.3 replaces the date and outage when all Alloy 800 sleeves shall be removed from service with a limitation on the individual sleeve service life from the date of installation. The allowed maximum service life previously approved for Alloy 800 sleeves remains unchanged.

The sleeve and portions of the installed sleeve/tube assembly that represent the reactor coolant pressure boundary will be monitored and a

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 13 of 15 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. Design criteria and design verification testing ensures that the margin of safety is not significantly different from the original SG tubes.

The proposed changes to TSs 3.4.20, 5.5.5, and 5.6.6 are changes consistent with TSTF-51 0, editorial corrections, and clarifications. The proposed changes will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above responses, FENOC concludes that the proposed amendment does not involve a 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 I Criteria 10 CFR 50.55a. Codes and Standards- (c) Reactor coolant pressure boundary.

Specifies that components which are part of the reactor coolant pressure boundary must meet the requirements for Class 1 components in Section Ill of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code).

10 CFR 50.55a further requires, in part, that throughout the service life of a pressurized water reactor facility, ASME Code Class 1 components meet the requirements, except design and access provisions and pre-service examination requirements, in Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," of the ASME Code, to the extent practical. This requirement includes the inspection and repair criteria of Section XI of the ASME Code. The design criteria of the Westinghouse leak-limiting sleeves were established to meet the loading condition and stress requirements of Section Ill of the ASME code (1995 edition, no agenda), which is consistent with the section of the ASME Code that applies to the original SG tubes.

10 CFR 50.65. Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants As a safety-related component relied upon to remain functional during and following a design basis event to ensure the integrity of the reactor coolant pressure boundary, under 10 CFR 50.65, licensees shall monitor SG performance against licensee-established goals.

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 14 of 15 With the proposed changes evaluated in section 3.0, the steam generators will continue to be monitored in accordance with this regulatory requirement.

10 CFR 50. Appendix A. General Design Criteria for Nuclear Power Plants General Design Criteria (GDC) 14, 15, 30, 31, and 32 of 10 CFR Part 50, Appendix A, define requirements for the reactor coolant pressure boundary with respect to structural and leakage integrity. Steam generator tubing and tube repairs constitute a major fraction of the reactor coolant pressure boundary surface area. Steam generator tubing and associated repair techniques and components, such as plugs and sleeves, must be capable of maintaining reactor coolant inventory and pressure.

The Steam Generator Program required by the proposed BVPS-1 and BVPS-2 Technical Specifications establishes performance criteria, repair criteria, repair methods, inspection periods and the methods necessary to meet them. These requirements provide reasonable assurance that tube integrity will be met in the interval between SG inspections.

The BVPS-1 construction permit was issued in June of 1970, before the GDC were published as Appendix A to 10 CFR 50 in July of 1971. Appendix 1A of the BVPS-1 Updated Final Safety Analysis Report (UFSAR) provides a discussion of the degree of conformance with the 1971 GDC. In Appendix 1A of the BVPS-1 UFSAR, it is noted that the BVPS-1 design conforms with the intent of GDC 14, 15, 30, 31, and 32.

10 CFR 50, Appendix B. Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.

10 CFR 50 Appendix B requires a quality assurance program for the design, fabrication, construction, and testing of structures, systems, and components in nuclear power plants. The requirements of Appendix B apply to all activities affecting the safety-related functions of those structures, systems, and components.

The activities include designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, refueling, and modifying safety-related structures, systems and components. The SGs and leak-limiting sleeves are considered safety-related components and, therefore, are required to meet the Appendix B requirements.

There are no proposed changes in this amendment request that impact this regulatory requirement.

4.3 Precedent The proposed changes to Technical Specification 5.5.5.2, "Unit 2 Steam Generator Program" that allow Westinghouse leak-limiting Alloy 800 sleeves to be used for five operating cycles is similar to the prior BVPS-2 License Amendment No. 170, which

Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit Nos. 1 and 2 Page 15 of 15 approved the initial use of Westinghouse leak-limiting Alloy 800 sleeves for five operating cycles (ADAMS Accession No. ML092590189).

The proposed changes toTS 3.4.20, "Steam Generator (SG) Tube Integrity," 5.5.5, "Steam Generator (SG) Program," and 5.6.6, "Steam Generator Tube Inspection Report" are consistent with sections 3.4.20, "Steam Generator (SG) Tube Integrity,"

5.5.9, "Steam Generator (SG) Program," and 5.6.7, "Steam Generator Tube Inspection Report" of TSTF-51 0, Revision 2 (ADAMS Accession No. ML110610350).

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

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 effluents 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.

Proposed Revision of Technical Specification (TS) 3.4.20, "Steam Generator (SG)

Tube Integrity"; TS 5.5.5, "Steam Generator (SG) Program"; and TS 5.6.6, "Steam Generator Tube Inspection Report" for the Beaver Valley Power Station, Unit Nos. 1 and 2 Attachment 1 Proposed Changes to Technical Specifications, Annotated Copy The following lists the Technical Specification pages included within Attachment 1:

3.4.20-1 3.4.20-2 5.5-4 5.5-5 5.5-6 5.5-7 5.5-8 5.5-9 5.5-10 5.5-11 5.5-12 5.5-13 5.6-4 5.6-5 5.6-6

SG Tube Integrity 3.4.20 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.20 Steam Generator (SG) Tube Integrity LCO 3.4.20 SG tube integrity shall be maintained.

All SG tubes satisfying the tube plugging or repair criteria shall be plugged or repaired( 1> in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

-NOTE-Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube plugging affected tube(s) is QLrepair criteria and not maintained until the next plugged or repaired( 1> in refueling outage or SG accordance with the Steam tube inspection.

Generator Program.

AND A.2 Plug or repair( 1> the Prior to entering affected tube(s) in MODE 4 following the accordance with the Steam next refueling outage Generator Program. or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.

( 1> SG Tube repair is only applicable to Unit 2.

Beaver Valley Units 1 and 2 3.4.20 - 1 Amendments 2:7-8-/ 494-

SG Tube Integrity 3.4.20 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.20.1 Verify SG tube integrity in accordance with the Steam In accordance Generator Program. with the Steam Generator Program SR 3.4.20.2 Verify that each inspected SG tube that satisfies the Prior to entering tube plugging or repair criteria is plugged or repaired< 1> MODE 4 following in accordance with the Steam Generator Program. a SG tube inspection

<1> SG Tube repair is only applicable to Unit 2.

Beaver Valley Units 1 and 2 3.4.20 - 2 Amendments 2:73-/ ~

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 lnservice Testing Program (continued)

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities,

c. The provisions of SR 3.0.3 are applicable to inservice testing activities, and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

5.5.5 Steam Generator (SG) Program A Steam Generator Program for Unit 1 and Unit 2 shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program for Unit 1 shall include the provisions of Specification 5.5.5.1 and the Steam Generator Program for Unit 2 shall include the provisions of Specification 5.5.5.2.

5.5.5.1 Unit 1 Steam GeneratorSG 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 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.

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 down1.--aRG all anticipated transients included in the design specification.t 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 Beaver Valley Units 1 and 2 5.5-4 Amendments 278/161

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

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

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

c. Provisions for SG Tube Repair Plugging 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.

d. Provisions for SG Tube Inspections Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube Fepaif plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. 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 installation.

Beaver Valley Units 1 and 2 5.5-5 Amendments 278 I 161

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

2. Inspect 100% of the tubes at sequential periods of 144, 108,72, and, thereafter, 60 effective full po\*Jer months. The first sequential period shall be considered to begin after tho first insorviso inspection of tho SGs. During eash period inspect 50% of tho tubes by the refueling outage nearest tho midpoint of tho period and tho remaining 50% by tho refueling outage nearest tho ond of the period. No SG shall operate for more than 72 effective full power months or three intervals between refueling outages (whichever is loss) witho1::1t being inspected.

After the first refueling outage following SG installation. inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections) . In addition. the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a. b. c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria. the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period . Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following SG installation. inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months. inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months. inspect 100% of the tubes. This constitutes the third inspection period ; and d) During the remaining life of the SGs. inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

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

Programs and Manuals 5.5 5.5 Programs and Manuals

3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected 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 results in more frequent inspections 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 GeneraterSG 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 downL-aAG all anticipated transients included in the design specificationJ 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, 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.

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

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

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 1x1 o-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 LeakageLEAKAGE."

c. Provisions for SG Tube Plugging or 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.

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 .

Beaver Valley Units 1 and 2 5.5-8 Amendments 278 I 161

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

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 or frepair11imit 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 or repaired, 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 5.5-9 Amendments 278 I 170

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 Steam Generator (SGt Program (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:

v V = SL MURL 1.0+NDE+Gr ( CL-.1.~

CL CL-6t)

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

VuRL = upper voltage repair limit VLRL = lower voltage repair limit VMuRL = mid-cycle upper voltage repair limit based on time into cycle VMLRL = mid-cycle lower voltage repair limit based on VMuRL and time into cycle

~t = 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- 10 Amendments 278/191 I

i II

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 or cold-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 or cold-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 2.22 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 or cold-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.

c) The F* methodology cannot be applied to the tubesheet region where a laser or TIG welded sleeve has been installed.

d. Provisions for SG Tube Inspections

-NOTE-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 plugging 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 plugging or 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, and d.5 and d.6 below, the inspection Beaver Valley Units 1 and 2 5.5 - 11 Amendments 278/172

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 Stear:n Generator (SGt Program (continued) 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 replacer:nent installation.

2. Inspect 100% of the tubes at sequential periods of 60 effective full po*Ner r:nonths. The first sequential period shall be considered to begin after the first in service inspection of the SGs. No SG shall operate for r:nore than 24 effective full power r:nonths or one interval betvteen refueling outages (whichever is less) without being inspected.

After the first refueling outage following SG installation, inspect each steam generator at least every 24 effective full power months or at least every refueling outage (whichever results in more frequent inspections).

In addition, inspect 100% of the tubes at sequential periods of 60 effective full power months beginning after the first refueling outage inspection following SG installation. Each 60 effective full power month inspection period may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation r:nechanisr:n that caused the crack indication shall not exceed 24 effective full power r:nonths or one interval between refueling outages (whichever is less). If definitive inforr:nation, such as frorn 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.

Beaver Valley Units 1 and 2 5.5- 12 Amendments 278 I 172

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

~-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).

e§.. 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.

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.

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

2. Westinghouse laser welded sleeves, WCAP-13483, Revision 2.

3. Westinghouse leak-limiting Alloy 800 sleeves, WCAP-15919-P, Revision 2. All Alloy 800 sleeves shall 13e FOFTlO'Jed from service 13y the spring of 2017 Unit 2 re~eling outage (2R1 Q) .An Alloy 800 sleeve shall remain in service for no more than five fuel cycles of operation starting from the outage when the sleeve was installed.

Beaver Valley Units 1 and 2 5.5- 13 Amendments 278 I 172 II II

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

The methodology listed in WCAP-14040-NP-A was used with two exceptions:

• ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limits for Section XI, Division 1."

• ASME,Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 1996 version.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.6.5 Post Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.6 Steam Generator (SG) Tube Inspection Report 5.6.6.1 Unit 1 SG Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.1, :unit 1 Steam Generator (SGt Program.: The report shall include:

a. The scope of inspections performed on each SG,

b. Aetive dDegradation 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 aGtive degradation mechanism,

f. The number and percentage of tubes plugged to date. and the effective plugging percentage in each steam generatorTotal number and pereentage of tubes plugged to date, and

g. The results of condition monitoring, including the results of tube pulls and in-situ testing,-aREt Beaver Valley Units 1 and 2 5.6 - 4 Amendments 291 I 178 i II 11

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Steam Generator (SG) Tube Inspection Report (continued)

h. The effective plugging peroentage for all plugging in eash SG.

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:t Program.:

The report shall include:

a. The scope of inspections performed on each SG,

b. Astive dDegradation 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 astive degradation mechanism,

f. The number and percentage of tubes plugged or repaired to date. and the effective plugging percentage in each steam generatorTotal number and persentage of tubes plugged or repaired to date,

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

h. The effestive plugging persentage for all plugging and tube repairs in eash SG, and

.b.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 GenoratorSG 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:

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

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

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

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.

As applicable, the report shall include the following hot-leg and cold-leg tubesheet region inspection results associated with the application ofF*:

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 / 172

Proposed Revision of Technical Specification (TS) 3.4.20, "Steam Generator (SG)

Tube Integrity"; TS 5.5.5, "Steam Generator (SG) Program"; and TS 5.6.6, "Steam Generator Tube Inspection Report" for the Beaver Valley Power Station, Unit Nos. 1 and 2 Attachment 2 Proposed Changes to Technical Specifications, Retyped Copy The following lists the Technical Specification pages included within Attachment 2:

3.4.20-1 3.4.20-2 5.5-4 5.5-5 5.5-6 5.5-7 5.5-8 5.5-9 5.5-10 5.5-11 5.5-12 5.6-4 5.6-5 5.6-6

SG Tube Integrity Retyped pages 3.4.20 provided for information 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.20 Steam Generator (SG) Tube Integrity LCO 3.4.20 SG tube integrity shall be maintained.

All SG tubes satisfying the tube plugging or repair criteria shall be plugged or repaired( 1> in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

-NOTE-Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A One or more SG tubes A1 Verify tube integrity of the 7 days satisfying the tube plugging affected tube(s) is or repair criteria and not maintained until the next plugged or repaired( 1> in refueling outage or SG accordance with the Steam tube inspection.

Generator Program.

AND A2 Plug or repair( 1> the Prior to entering affected tube(s) in MODE 4 following the accordance with the Steam next refueling outage Generator Program. or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.

(1) SG Tube repair is only applicable to Unit 2.

Beaver Valley Units 1 and 2 3.4.20 - 1 Amendments I I

li

,I II

SG Tube Integrity Retyped pages 3.4.20 provided for information SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.20.1 Verify SG tube integrity in accordance with the Steam In accordance Generator Program. with the Steam Generator Program SR 3.4.20.2 Verify that each inspected SG tube that satisfies the Prior to entering tube plugging or repair criteria is plugged or repaired< 1> MODE 4 following in accordance with the Steam Generator Program. a SG tube inspection

<1> SG Tube repair is only applicable to Unit 2.

Beaver Valley Units 1 and 2 3.4.20-2 Amendments I

Programs and Manuals Retyped pages provided for information 5.5 5.5 Programs and Manuals 5.5.4 lnservice Testing Program (continued)

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities,

c. The provisions of SR 3.0.3 are applicable to inservice testing activities, and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

5.5.5 Steam Generator (SG) Program A Steam Generator Program for Unit 1 and Unit 2 shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program for Unit 1 shall include the provisions of Specification 5.5.5.1 and the Steam Generator Program for Unit 2 shall include the provisions of Specification 5.5.5.2.

5.5.5.1 Unit 1 SG 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 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.

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), 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 Beaver Valley Units 1 and 2 5.5-4 Amendments

Programs and Manuals Retyped pages 5.5 provided for information 5.5 Programs and Manuals 5.5.5.1 Unit 1 SG Program (continued) 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.

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

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

c. Provisions for SG Tube Plugging 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.

d. Provisions for SG Tube Inspections Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. 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 installation.

2. After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the I

i II Beaver Valley Units 1 and 2 5.5-5 Amendments II jl

Retyped pages Programs and Manuals provided for information 5.5 5.5 Programs and Manuals 5.5.5.1 Unit 1 SG Program (continued) number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months.

This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). 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 Beaver Valley Units 1 and 2 5.5-6 Amendments

Programs and Manuals Retyped pages 5.5 provided for information 5.5 Programs and Manuals 5.5.5.2 Unit 2 SG 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), 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, 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 1x1 o-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 Beaver Valley Units 1 and 2 5.5-7 Amendments

Retyped pages Programs and Manuals 5.5 provided for information 5.5 Programs and Manuals 5.5.5.2 Unit 2 SG Program (continued)

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 Plugging or 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.

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 or 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 plugged or repaired, except as noted in 5.5.5.2.c.4.c below.

Beaver Valley Units 1 and 2 5.5-8 Amendments

Programs and Manuals Retyped pages provided for information 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 SG Program (continued) 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.

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:

V =--------V=S~L~~~-

MURL 1.0+NDE+Gr ( CL-L'lt)

CL CL-L'lt)

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

VuRL = upper voltage repair limit VLRL = lower voltage repair limit VMuRL = mid-cycle upper voltage repair limit based on time into cycle VMLRL = mid-cycle lower voltage repair limit based on VMuRL and time into cycle

~t = 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 Beaver Valley Units 1 and 2 5.5-9 Amendments II I

Retyped pages Programs and Manuals provided for information 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 SG Program (continued)

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.

5. The F* methodology, as described below, may be applied to the expanded portion of the tube in the hot-leg or cold-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 or cold-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 2.22 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 or cold-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.

c) The F* methodology cannot be applied to the tubesheet region where a laser or TIG welded sleeve has been installed.

d. Provisions for SG Tube Inspections

-NOTE-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 plugging 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 Beaver Valley Units 1 and 2 5.5- 10 Amendments

Programs and Manuals Retyped pages provided for information 5.5 5.5 Programs and Manuals 5.5.5.2 Unit 2 SG Program (continued) 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 plugging or 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 and d.5 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.

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

2. After the first refueling outage following SG installation, inspect each steam generator at least every 24 effective full power months or at least every refueling outage (whichever results in more frequent inspections).

In addition, inspect 100% of the tubes at sequential periods of 60 effective full power months beginning after the first refueling outage inspection following SG installation. Each 60 effective full power month inspection period may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period.

Beaver Valley Units 1 and 2 5.5- 11 Amendments I

II

Programs and Manuals Retyped pages 5.5 provided for information 5.5 Programs and Manuals 5.5.5.2 Unit 2 SG Program (continued)

3. 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.

4. 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).

5. 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.

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.

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

2. Westinghouse laser welded sleeves, WCAP-13483, Revision 2.

3. Westinghouse leak-limiting Alloy 800 sleeves, WCAP-15919-P, Revision 2. An Alloy 800 sleeve shall remain in service for no more than five fuel cycles of operation starting from the outage when the sleeve was installed.

Beaver Valley Units 1 and 2 5.5- 12 Amendments I

11 I!

Retyped pages Reporting Requirements provided for information 5.6 5.6 Reporting Requirements 5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

The methodology listed in WCAP-14040-NP-A was used with two exceptions:

• ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limits for Section XI, Division 1."

• ASME,Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 1996 version.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.6.5 Post Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.6 Steam Generator (SG) Tube Inspection Report 5.6.6.1 Unit 1 SG Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.1, "Unit 1 SG Program." The report shall include:

a. The scope of inspections performed on each SG,

b. 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 degradation mechanism,

f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator, and

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

Beaver Valley Units 1 and 2 5.6-4 Amendments

Retyped pages Reporting Requirements provided for information 5.6 5.6 Reporting Requirements 5.6.6 Steam Generator (SG) 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 SG Program." The report shall include:

a. The scope of inspections performed on each SG,

b. 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 degradation mechanism,

f. The number and percentage of tubes plugged or repaired to date, and the effective plugging percentage in each steam generator,

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

h. 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 SG 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.

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

Beaver Valley Units 1 and 2 5.6-5 Amendments

Reporting Requirements Retyped pages 5.6 provided for information 5.6 Reporting Requirements 5.6.6.2 Unit 2 SG Tube Inspection Report (continued)

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.

As applicable, the report shall include the following hot-leg and cold-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 II II

Proposed Revision of Technical Specification (TS) 3.4.20, "Steam Generator (SG)

Tube Integrity"; TS 5.5.5, "Steam Generator (SG) Program"; and TS 5.6.6, "Steam Generator Tube Inspection Report" for the Beaver Valley Power Station, Unit Nos. 1 and 2 Attachment 3 Proposed Changes to Technical Specification Bases, Annotated Copy The following lists the Technical Specification Bases pages included within Attachment 3:

8 3.4.20-1 8 3.4.20-2 8 3.4.20-3 B 3.4.20-4 B 3.4.20-5 8 3.4.20-6 B 3.4.20-7 8 3.4.20-8

For Information Only SG Tube Integrity B 3.4.20 No changes to this page.

Provided for context.

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.20 Steam Generator (SG) Tube Integrity BASES BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers.

The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.4, "RCS Loops- MODES 1 and 2," LCO 3.4.5, "RCS Loops- MODE 3," LCO 3.4.6, "RCS Loops- MODE 4," and LCO 3.4.7, "RCS Loops- MODE 5, Loops Filled."

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

Steam generator tubing is subject to a variety of degradation mechanisms. Depending upon materials and design, steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG pertormance criteria are used to manage SG tube degradation.

Specification 5.5.5, "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 5.5.5, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG pertormance criteria are described in Specification 5.5.5. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).

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BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes that following reactor trip the contaminated secondary fluid is released to the atmosphere via safety valves.

Environmental releases before reactor trip are discharged through the main condenser.

For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity," limits. Pre-accident and concurrent iodine spikes are assumed in accordance with applicable regulatory guidance. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of 10 CFR 50.67 (Ref. 2) as supplemented by Regulatory Guide 1.183 (Ref. 3) and within GDC-19 (Ref. 4) values.

Unit 1:

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is conservatively assumed to include the total primary to secondary LEAKAGE from all SGs of 450 gpd (i.e., 150 gpd per steam generator) or is assumed to increase to 450 gpd as a result of accident induced conditions.

Currently, the Unit 1 safety analyses do not specifically assume additional primary to secondary LEAKAGE due to accident induced conditions.

Unit 2:

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses, the steam discharge to the atmosphere is conservatively assumed to include the total primary to secondary LEAKAGE from all SGs of 450 gpd (i.e., 150 gpd per steam generator) or is assumed to increase to 450 gpd as a result of accident induced conditions for all accidents other than the Unit 2 main steam line break (MSLB).

Currently, the Unit 2 MSLB safety analysis is the only analysis that specifically assumes additional primary to secondary LEAKAGE due to accident induced conditions.

For the Unit 2 main steam line break (MSLB) analysis, an increased leakage assumption is applied. In support of voltage based repair criteria pursuant to Generic Letter 95-05 (Ref. 5) analyses were performed to Beaver Valley Units 1 and 2 B 3.4.20-2 Revision 0

SG Tube Integrity For Information Only B 3.4.20 BASES APPLICABLE SAFETY ANALYSES (continued) determine the maximum MSLB induced primary to secondary leak rate that could occur without offsite doses exceeding the limits of 10 CFR 50.67 (Ref. 2) as supplemented by Regulatory Guide 1.183 (Ref. 3) and without control room doses exceeding GDC-19 (Ref. 4). An additional 2.1 gpm leakage is assumed in the Unit 2 MSLB analysis resulting from accident conditions. Therefore, in the MSLB analysis, the steam discharge to the atmosphere includes primary to secondary LEAKAGE equivalent to the operational leakage limit of 150 gpd per SG and an additional 2.1 gpm which results in a total assumed accident induced leakage of 2.4 gpm.

The combined projected leak rate from all sources (i.e., voltage based repair criteria, application ofF*, freespan crack, leaking plug, leakage past sleeves, etc.) for each SG must be less than the maximum allowable steam line break leak rate limit in any one steam generator (i.e., 2.2 gpm) in order to maintain a total assumed accident induced leakage of:::;; 2.4 gpm as explained above. Maintaining the total assumed accident induced leakage to:::;; 2.4 gpm limits the resulting dose to within the requirements of 10 CFR 50.67 (Ref. 2) as supplemented by Regulatory Guide 1.183 (Ref. 3) and within GDC-19 (Ref. 4) values during a postulated steam line break event.

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36( c)(2)(ii).

LCO A Note modifies the LCO to indicate that any reference to the repair of SG tubes is only applicable to Unit 2 at this time. The Unit 1 "Steam Generator Program" (in Specification 5.5.5) has no provision for SG tube repair.

The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the plugging or repair criteria be plugged or repaired in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging or repair criteria is repaired or removed from service by plugging. If a tube was determined to satisfy the plugging or repair criteria but was not plugged or repaired , the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall and any repairs made to it, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria.

The SG performance criteria are defined in Specification 5.5.5, "Steam Beaver Valley Units 1 and 2 B 3.4.20-3 Revision G

SG Tube Integrity For Information Only 8 3.4.20 No changes to this page.

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BASES LCO (continued)

Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting bursUcollapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code, Section Ill, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code, Section Ill, Subsection NB (Ref. 6) and Draft Regulatory Guide 1.121 (Ref. 7).

The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions as described in the Applicable Safety Analyses section of this Bases. The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

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SG Tube Integrity For Information Only B 3.4.20 BASES LCO (continued)

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4. 13, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

ACTIONS The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated Required Actions.

A.1 and A.2 A Note modifies Condition A and Required Action A.2 to indicate that any reference to the repair of SG tubes is only applicable to Unit 2 at this time.

The Unit 1 "Steam Generator Program" (in Specification 5.5.5) has no provision for SG tube repair.

Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube plugging or repair criteria but were not plugged or repaired in accordance with the Steam Generator Program as required by SR 3.4.20.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG plugging or repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube Beaver Valley Units 1 and 2 B 3.4.20-5 Revision G

SG Tube Integrity For Information Only B 3.4.20 BASES ACTIONS (continued) that should have been plugged or repaired has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Condition B applies.

A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged or repaired prior to entering MODE 4 following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.20.1 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

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For Information Only SG Tube Integrity B 3.4.20 BASES SURVEILLANCE REQUIREMENTS (continued)

The Steam Generator Program in conjunction with the degradation assessment determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube plugging or repair criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program and the degradation assessment also specify the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 3.4.20.1.

The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 8). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.5.5 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 5.5.5 until subsequent inspections support extending the inspection interval.

SR 3.4.20.2 A Note modifies SR 3.4.20.2 to indicate that any reference to the repair of SG tubes is only applicable to Unit 2 at this time. The Unit 1 "Steam Generator Program" (in Specification 5.5.5) has no provision for SG tube repair.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging or repair criteria is repaired or removed from service by plugging. The tube plugging or repair criteria delineated in Specification 5.5.5 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube plugging or repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s).

Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

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For Information Only SG Tube Integrity B 3.4.20 BASES SURVEILLANCE REQUIREMENTS (continued)

Steam generator tube repairs are only performed using approved repair methods as described in the Steam Generator Program (Specification 5.5.5).

The Frequency of prior to entering MODE 4 following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the plugging or repair criteria are plugged or repaired prior to subjecting the SG tubes to significant primary to secondary pressure differential.

REFERENCES 1. NEI 97-06, "Steam Generator Program Guidelines."

2. 10 CFR 50.67, Accident Source Term.

3. Regulatory Guide 1.183, "Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Reactors."

4. 10 CFR 50 Appendix A, GDC 19.

5. NRC Generic Letter 95-05, "Voltage-Based Repair Criteria For Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking."

6. ASME Boiler and Pressure Vessel Code, Section Ill, Subsection NB.

7. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

8. EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."

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