ML14078A037

License Amendment Request (LAR) for Adoption of Technical Specification Task Force (TSTF) Change Travelers TSTF-479 and TSTF-497

ML14078A037
Person / Time
Site:	Oconee
Issue date:	03/14/2014
From:	Batson S L Duke Energy Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR 2013-04, ONS-2014-015
Download: ML14078A037 (49)

Text

DUKE Scoff L. Baton Vice President ENERGY, Oconee Nuclear Station Duke Energy ONO1VP 1 7800 Rochester Hwy Seneca, SC 29672 ONS-2014-015 o: 864.873.3274 10 CFR 50.90 8 864.873.4208 March 14, 2014 Scott.Batson@duke-energy.com ATTN: Document Control Desk U. S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852 Duke Energy Carolinas, LLC (Duke Energy)Oconee Nuclear Station, Units 1, 2, and 3 Docket Nos. 50-269, 50-270, and 50-287 Renewed License Nos. DPR-38, DPR-47, and DPR-55

Subject:

License Amendment Request (LAR) for Adoption of Technical Specification Task Force (TSTF) Change Travelers TSTF-479 and TSTF-497 Oconee Nuclear Station (ONS) LAR No. 2013-04 In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Duke Energy is submitting a request for an amendment to the Technical Specifications (TS) for Oconee Nuclear Station (ONS), Units 1, 2, and 3. The proposed amendment would revise the TS Administrative Controls Inservice Testing Program (i.e., TS 5.5.9) and references in the TS Bases to reflect the current edition of the American Society of Mechanical Engineers (ASME) Code that is referenced in 10 CFR 50.55a(b).

U. S. Nuclear Regulatory Commission (NRC) regulations require the Inservice Testing Program at nuclear power plants to be revised every 120 months to comply with the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b).

The adoption of the latest edition renders incorrect certain statements in the TS Administrative Controls Inservice Testing Program and in the TS Bases. These changes are consistent with NRC-approved Revision 0 to TSTF Improved Standard Technical Specification Change Travelers TSTF-479, "Changes to Reflect Revision of 10 CFR 50.55a," and Revision 0 to TSTF-497, "Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less." The enclosure to this letter provides an evaluation of the proposed TS changes. Regulatory analysis (including the No Significant Hazards Consideration) and environmental considerations are provided in Sections 5 and 6 of the enclosure, respectively.

Attachments 1 and 2 provide mark-ups of the corrected TS and TS Bases pages, respectively.

Attachments 3 and 4 provide retyped (clean)TS and TS Bases pages, respectively.

Once this amendment request is approved, the amendment will be implemented within 120 days. Duke Energy will also update applicable sections of the ONS Updated Final Safety Analysis Report (UFSAR), as necessary, and submit the updated UFSAR sections in accordance with 10 CFR 50.71(e).

There are no new regulatory commitments being made as a result of the proposed change.X06c 1 U. S. Nuclear Regulatory Commission March 14, 2014 Page 2 If there are any questions regarding the content of this document or if additional information is needed, please contact Sandra Severance, Regulatory Affairs Group, Oconee Nuclear Station, at (864) 873-3466.I declare under penalty of perjury that the foregoing is correct and true. Executed on the 14th day of March, 2014.Sincerely, Scott L. Batson Site Vice President Oconee Nuclear Station

Enclosure:

Evaluation of the Proposed Changes Attachments:

1. Attachment 1 -Markups of Technical Specification Pages 2. Attachment 2 -Markups of Technical Specification Bases Pages 3. Attachment 3 -Revised Technical Specification Pages 4. Attachment 4 -Revised Technical Specification Bases Pages U. S. Nuclear Regulatory Commission March 14, 2014 Page 3 cc w/enclosure and attachments:

Mr. Victor McCree Administrator, Region II U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Mr. Richard Guzman Senior Project Manager (by electronic mail only)U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation 11555 Rockville Pike Mail Stop O-8C2 Rockville, MD 20852 Mr. Eddy Crowe NRC Senior Resident Inspector Oconee Nuclear Station Ms. Susan E. Jenkins, Manager, Infectious and Radioactive Waste Management, Division of Waste Management South Carolina Department of Health & Environmental Control 2600 Bull Street, Columbia, SC 29201 ENCLOSURE EVALUATION OF THE PROPOSED CHANGES LICENSE AMENDMENT REQUEST NO. 2013-04

Subject:

License Amendment Request for the Adoption of Technical Specification Task Force (TSTF) Change Travelers TSTF-479 and TSTF-497 1

SUMMARY

DESCRIPTION 2 BACKGROUND 3 DESCRIPTION OF PROPOSED CHANGES 4 TECHNICAL ANALYSIS 5 REGULATORY ANALYSIS 5.1 No Significant Hazards Consideration

5.2 Applicable

Regulatory Requirements/Criteria

5.3 Precedence

6 ENVIRONMENTAL CONSIDERATION 7 REFERENCES Enclosure

-Evaluation of Proposed Changes License Amendment Request No. 2013-04 Page 1 of 9 1

SUMMARY

DESCRIPTION The proposed amendment would revise the Oconee Nuclear Station (ONS) Units 1, 2, and 3 Technical Specifications (TS) Administrative Controls Inservice Testing Program (i.e., TS 5.5.9) and references in the TS Bases (TSB) to reflect the current edition of the American Society of Mechanical Engineers (ASME) Code that is referenced in 10 CFR 50.55a(b).

U. S. Nuclear Regulatory Commission (NRC) regulations require the Inservice Testing Program (ITP) at nuclear power plants to be revised every 120 months to comply with the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b).

The adoption of the latest edition renders select statements in the TS Administrative Controls ITP description and in the TSB incorrect.

These changes are consistent with NRC-approved Technical Specification Task Force (TSTF) Improved Standard Technical Specification Change Travelers TSTF-479, "Changes to Reflect Revision of 10 CFR 50.55a," (Ref. 1) and TSTF-497, "Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less" (Ref. 2).In addition, the proposed amendment corrects an identified typographical error in TS 5.5.8,"Reactor Coolant Pump Flywheel Inspection Program." A detailed description of the proposed changes is provided in Section 3. A technical analysis of the proposed changes is provided in Section 4. The marked-up TS and TS Bases pages associated with this license amendment request (LAR) are provided in Attachments 1 and 2, respectively, and the retyped (clean) TS and TSB pages are provided in Attachments 3 and 4, respectively.

Once this LAR is approved, the amendment will be implemented within 120 days. There are no new regulatory commitments being made as a result of this proposed change.2 BACKGROUND In 1990, the ASME published the initial edition of the ASME Operation and Maintenance (OM) Code which gives rules for inservice testing of pumps and valves at nuclear power plants. The ASME intended that the ASME OM Code replace Section XI of the ASME Boiler and Pressure Vessel (B&PV) Code for inservice testing of pumps and valves. The 1995 edition of the ASME OM Code was incorporated by reference into 10 CFR 50.55a on September 22, 1999 (Ref. 3). Since 10 CFR 50.55a(f)(4)(ii) requires that inservice testing comply with the requirements of the latest edition and addenda of the ASME Code incorporated into 10 CFR 50.55a(b), TS 5.5.9 must be revised to reference the ASME OM Code. TSTF-479 was developed to provide licensees a standard method to request NRC approval of this required TS revision.

The NRC approved TSTF-479, as an administrative change to the Improved Standard Technical Specifications (ISTS) NUREGs, in a letter dated December 6, 2005 (Ref. 4).Although the NRC approved TSTF-479, the NRC expressed concerns with the TSTF changes to paragraph b of the ITP TS in a February 23, 2006 meeting with the TSTF Group.Specifically, the NRC felt that TSTF-479 did not provide adequate justification for applying Surveillance Requirement (SR) 3.0.2 to Frequencies specified in the ITP as greater than two years. Thus, the TSTF Group developed TSTF-497 to revise paragraph b of the ITP TS to Enclosure

-Evaluation of Proposed Changes License Amendment Request No. 2013-04 Page 2 of 9 specify the provisions of SR 3.0.2 are applicable to Inservice Testing Frequencies specified as two years or less. The NRC approved TSTF-497 in a letter dated October 4, 2006 (Ref.5).3 DESCRIPTION OF PROPOSED CHANGES Duke Energy proposes to modify the TS and TSB (for information only). The proposed change to ONS TS 5.5.8 only corrects a typographical error. The proposed changes to ONS TS 5.5.9, and in the TSB, will replace reference to ASME B&PV Code with reference to ASME OM Code for pump and valve testing only. The TS 5.5.9 proposed changes adopt changes specified in NRC-approved TSTF-479 and TSTF-497 without variations or deviations.

The detailed proposed changes are listed below.For TS 5.5.8:* In the third sentence, change the word "urface" to "surface." For TS 5.5.9: " TS 5.5.9a -Replace "specified in Section XI of the ASME Boiler and Pressure Vessel Code" with "applicable to the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code)."" TS 5.5.9a -In the column heading that states "ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice testing activities," change"Boiler and Pressure Vessel" to "OM."" TS 5.5.9b -Between the words "Frequencies" and "for," add new text as follows: "and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program."* TS 5.5.9d -Change "Boiler and Pressure Vessel" to "OM." For TSB B 3.4.10:* LCO Section -In the fifth line, change the term "... per ASME Section XI requirements..

." to ".... per ASME Code requirements..."" Surveillance Requirements Section -In the first paragraph, delete "of Section XI" in the second line." References Section -Change Reference 2 to state "ASME Code for Operation and Maintenance of Nuclear Power Plants." For TSB B 3.4.14:* References Section -Change Reference 7 to state "ASME Code for Operation and Maintenance of Nuclear Power Plants."

Enclosure

-Evaluation of Proposed Changes License Amendment Request No. 2013-04 Page 3 of 9 For TSB B 3.5.2: " Surveillance Requirements Section -In SR 3.5.2.3, third line, delete "Section XI of."* Surveillance Requirements Section -In SR 3.5.2.3, second sentence, revise wording to state" "SRs are specified in the Inservice Testing Program of the ASME Code."* References Section -Change Reference 5 to state "ASME Code for Operation and Maintenance of Nuclear Power Plants." For TSB B 3.5.3:* Surveillance Requirements Section -In SR 3.5.3.3, third line, delete "Section XI of."* Surveillance Requirements Section -In SR 3.5.3.3, second sentence, revise wording to state" "SRs are specified in the Inservice Testing Program of the ASME Code."" References Section -Change Reference 6 to state "ASME Code for Operation and Maintenance of Nuclear Power Plants." For TSB B 3.6.5: " Surveillance Requirements Section -In SR 3.6.5.3, fifth line, delete "Section X1 of."" References Section -Change Reference 4 to state "ASME Code for Operation and Maintenance of Nuclear Power Plants." For TSB B 3.7.1: " Surveillance Requirements Section -In SR 3.7.1.1, first paragraph, third line, change"ANSI/ASME" to "ASME."" Surveillance Requirements Section -In SR 3.7.1.1, second paragraph, first line, change "ANSI/ASME Standard" to "ASME Code."* References Section -Change Reference 6 to state "ASME Code for Operation and Maintenance of Nuclear Power Plants." For TSB B 3.7.3: " Surveillance Requirements Section -In SR 3.7.3.1, second paragraph, sixth line, delete ", Section Xl."* References Section -Change Reference 2 to state "ASME Code for Operation and Maintenance of Nuclear Power Plants." For TSB B 3.7.5: " Surveillance Requirements Section -In SR 3.7.5.2, fifth line, delete "Section XI of."" Surveillance Requirements Section -In SR 3.7.5.2, second paragraph, third line, delete ", Section Xl."" References Section -Change Reference 3 to state "ASME Code for Operation and Maintenance of Nuclear Power Plants."

Enclosure

-Evaluation of Proposed Changes License Amendment Request No. 2013-04 Page 4 of 9 The above TS and TSB changes are identified in Attachments 1 and 2, respectively.

4 TECHNICAL ANALYSIS The purposes of the ONS ITP are to assess the operational readiness of pumps and valves, to detect degradation that might affect component OPERABILITY, and to maintain safety margins with provisions for increased surveillance and corrective action. NRC regulation 10 CFR 50.55a defines the requirements for applying industry codes to each licensed nuclear powered facility.

Licensees are required by 10 CFR 50.55a(f)(4)(i) to initially prepare programs to perform inservice testing of certain ASME Section III, Code Class 1, 2, and 3 pumps and valves during the initial 120-month interval of unit operation.

NRC regulation 10 CFR 50.55a(f)(4)(ii) requires that the ITP be revised during successive 120-month intervals of unit operation to comply with the latest edition and addenda of the Code incorporated by reference in paragraph (b) 12 months prior to the start of the interval.Section Xl of the ASME Codes has been revised on a continuing basis over the years to provide updated requirements for the inservice inspection and inservice testing of components.

Until 1990, the ASME Code requirements addressing the inservice testing of pumps and valves were contained in Section Xl, Subsections IWP (for pumps) and IWV (for valves). In 1990, the ASME published the initial edition of the OM Code that provides the rules for inservice testing of pumps and valves. Since the establishment of the 1990 Edition of the OM Code, the rules for inservice testing of pumps are no longer being updated in Section XI. As identified in NRC SECY-99-017, dated January 13, 1999 (Ref. 6), the NRC has generally considered the evolution of the ASME Code to result in a net improvement in the measures for inspecting piping and components and for testing pumps and valves.The TS ITP is revised to indicate that the provisions of SR 3.0.2 are applicable to other Inservice Testing Frequencies, of two years or less, that are not specified in the ITP. The ITP may have Frequencies for testing that are based on risk and do not conform to the standard testing Frequencies specified in the TS. For example, an ITP may use ASME Code Case OMN-1, "Alternative Rules for Preservice and Inservice Testing of Certain Electric Motor-Operated Valve Assemblies in Light-Water Reactor Plants," in lieu of stroke time testing. The Frequency of the Surveillance may be determined through a mix of risk informed and performance based means in accordance with the ITP. This is consistent with the guidance in NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants" (Ref. 7), which indicates that the 25% extension of the interval specified in the Frequency would apply to increased frequencies the same way that it applies to regular frequencies.

If a test interval is specified in 10 CFR 50.55a, the TS SR 3.0.2 Bases indicates that the requirements of the regulation take precedence over the TS.

Enclosure

-Evaluation of Proposed Changes License Amendment Request No. 2013-04 Page 5 of 9 5 REGULATORY ANALYSIS 5.1 NO SIGNIFICANT HAZARDS CONSIDERATION Duke Energy Carolinas, LLC (Duke Energy), has evaluated the proposed changes to the Oconee Nuclear Station (ONS) Technical Specifications (TS) using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

An analysis of the issue of no significant hazards consideration is presented below: Description of Amendment Reauest The proposed amendment would correct a typographical error in TS 5.5.8, "Reactor Coolant Pump Flywheel Inspection Program," and revise TS 5.5.9, "Inservice Testing Program," to include testing frequencies applicable to the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code instead of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI. Additionally, TS 5.5.9 would also be revised to indicate that there may be some non-standard Frequencies utilized in the Inservice Testing Program in which provisions of SR 3.0.2 are applicable.

As described below, Duke Energy concludes that the change does not meet any of the three criteria for a significant hazards consideration.

Basis for Proposed No Significant Hazards Consideration Determination As required by 10 CFR 50.91 (a), the Duke Energy analysis of the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is presented below: 1. Does the Proposed Change Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated?

Response:

No The proposed change corrects a typographical error in TS 5.5.8, "Reactor Coolant Pump Flywheel Inspection Program," and revises TS 5.5.9, "Inservice Testing Program," for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves which are classified as ASME Code Class 1, Class 2 and Class 3. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves.The proposed change does not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. The proposed change does not involve the addition or removal of any equipment, or any design changes to the facility.

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

Enclosure

-Evaluation of Proposed Changes License Amendment Request No. 2013-04 Page 6 of 9 2. Does the Proposed Change Create the Possibility of a New or Different Kind of Accident from any Accident Previously Evaluated?

Response:

No The proposed change corrects a typographical error in TS 5.5.8, "Reactor Coolant Pump Flywheel Inspection Program," and revises TS 5.5.9, "Inservice Testing Program," for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves which are classified as ASME Code Class 1, Class 2 and Class 3. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves.The proposed change does not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed), nor does it involve a change in the methods governing normal plant operation.

The proposed change will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism.

Additionally, there is no change in the types or increases in the amounts of any effluent that may be released offsite and there is no increase in individual or cumulative occupational exposure.

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

3. Does the Proposed Change Involve a Significant Reduction in a Margin of Safety?Response:

No The proposed change corrects a typographical error in TS 5.5.8, "Reactor Coolant Pump Flywheel Inspection Program," and revises TS 5.5.9, "Inservice Testing Program," for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves which are classified as ASME Code Class 1, Class 2 and Class 3. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves. The safety function of the affected pumps and valves will be maintained.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.Based upon the above analysis, Duke Energy concludes that the requested change does not involve a significant hazards consideration, as set forth in 10 CFR 50.92(c),"Issuance of Amendment."

Enclosure

-Evaluation of Proposed Changes License Amendment Request No. 2013-04 Page 7 of 9 5.2 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA NRC regulation 10 CFR 50.55a defines the requirements for applying industry codes to each licensed nuclear powered facility.

Licensees are required by 10 CFR 50.55a(f)(4)(i) to initially prepare programs to perform inservice testing of certain ASME Section III, Code Class 1, 2, and 3 pumps and valves during the initial 120-month interval.

The regulations require that programs be developed utilizing the latest edition and addenda incorporated into paragraph (b) of 10 CFR 50.55a on the date 12 months prior to the date of issuance of the operating license subject to the limitations and modification identified in paragraph (b).The proposed changes do not:* Alter the design or function of any system;* Result in any changes in the qualifications of any component; or* Result in the reclassification of any component's status in the areas of shared, safety-related, independent, redundant, and physically or electrically separated.

In addition, this Technical Specification change will not reduce the leak-tightness of the containment.

As such, there are no changes being proposed such that compliance with the regulatory requirements of 10 CFR 50.55a would not be fulfilled.

Therefore, 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) Issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.5.3 PRECEDENCE A review of the NRC's Agencywide Documents Access and Management System (ADAMS) for prior TSTF-479/-497 license amendments issued by the NRC to nuclear power plants resulted in the following documents of precedence.

• NRC Letter to Nuclear Management Company, LLC, "Prairie Island Nuclear Generating Plant, Units 1 and 2 -Issuance of Amendments Re: Incorporation of Technical Specification Task Force Travelers TSTF-479, TSTF-485 and TSTF-497 (TAC Nos. MD5983 and MD5984)," dated June 27, 2008 [ADAMS Accession No. ML081650272]." NRC Letter to Exelon Generation Company, LLC, "Braidwood Station, Units 1 and 2; Byron Station, Units Nos. 1 and 2; Dresden Nuclear Power Station, Units 2 and 3; Limerick Generating Station, Units 1 and 2; Oyster Creek Nuclear Generating Station; Peach Bottom Atomic Power Station, Units 2 and 3; Quad Cities Nuclear Power Station, Units 1 and 2; and Three Mile Island Nuclear Station, Unit 1 -Issuance of Amendments That Adopt Technical Specification Task Force (TSTF) Change Traveler TSTF-479 and TSTF-497 (TAC Nos.MD6530 Thru MD6543)," dated August 28, 2008 [ADAMS Accession No.ML080600330].

Enclosure

-Evaluation of Proposed Changes License Amendment Request No. 2013-04 Page 8 of 9* NRC Letter to Duke Energy Carolinas, LLC, "McGuire Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Request To Revise Technical Specification 5.5.8, 'Inservice Testing Program,'

To Adopt Technical Specification Change Travelers TSTF-479, Rev. 0 and TSTF-497, Rev. 0 (TAC Nos. MD9581 and MD9582)," dated August 17, 2009 [ADAMS Accession No. ML092240085].

• NRC Letter to Duke Energy Carolinas, LLC, "Catawba Nuclear Station, Units 1 and 2, Issuance of Amendments Adopting TSTF-479, Revision 0, and TSTF-497, Revision 0 (TAC Nos. MD9965 and MD9966)," dated October 30, 2009[ADAMS Accession No. ML092380588]." NRC Letter to Entergy Operations, Inc., "Arkansas Nuclear One, Unit No. 1 -Issuance Of Amendment Re: Modifications To Technical Specifications To Reflect Revision To 10 CFR 50.55a, Technical Specification Task Force Change Travelers TSTF-479-A and TSTF-497-A (TAC No. ME1829)," dated December 23, 2009 [ADAMS Accession No. ML093060132]." NRC Letter to Entergy Operations, Inc., "Arkansas Nuclear One, Unit No. 2 -Issuance of Amendment Re: Modifications To Technical Specifications To Reflect Adoption of Technical Specification Task Force (TSTF) Change Travelers TSTF-479-A and TSTF-497-A (TAC No. ME4118)," dated November 5, 2010[ADAMS Accession No. MLI 02010520].

• NRC Letter to Arizona Public Service Co., "Palo Verde Nuclear Generating Station, Units 1, 2, and 3 -Issuance of Amendments Re: Revise Technical Specification 5.5.8, Inservice Testing Program (TAC Nos. ME3914, ME3915, and ME3916)," dated January 19, 2011 [ADAMS Accession No. MLI103560088].

6 ENVIRONMENTAL CONSIDERATION The proposed change would modify requirements with respect to testing of facility components located within the restricted area, as defined in 10 CFR 20, or would change a surveillance requirements only to the extent of the ASME Code referenced during surveillance performance.

However, the proposed change 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 change 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 change.7 REFERENCES

1. TSTF Letter TSTF-04-15, dated December 2, 2004, "TSTF-479, Revision 0, 'Changes to Reflect Revision of CFR 50.55a"' [ADAMS Accession No. ML052990317].

2. TSTF Letter TSTF-06-14, dated July 12, 2006, "TSTF-497, Revision 0, 'Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less"' [ADAMS Accession No. ML061930221].

Enclosure

-Evaluation of Proposed Changes License Amendment Request No. 2013-04 Page 9 of 9 3. Federal Register Notice 64 FR 51370, dated September 22, 1999, "Industry Codes and Standards; Amended Requirements." 4. NRC Letter to Technical Specifications Task Force, dated December 6, 2005, "Status of TSTF 343, 479, 482, 485" [ADAMS Accession No. ML053460302].

5. NRC Letter to Technical Specifications Task Force, dated October 4, 2006, Approving TSTF-497, Revision 0 [ADAMS Accession No. ML062780321].

6. NRC SECY-99-017, dated January 13, 1999, "Proposed Amendment to 10 CFR 50.55a." 7. NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants," Revision 2 (Draft Report for Comment), August 2011.

License Amendment Request No. 2013-04 ATTACHMENT I Markups of Technical Specification Pages[2 pages following this cover page]NOTE: Attached are markups of existing TS Pages 5.0-12 and -13 which incorporate the changes described in the Letter Enclosure.

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.7 Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity.

The program shall include baseline measurements prior to initial operations.

The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, as amended by relief granted in accordance with 10 CFR 50.55a(a)(3).

The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

5.5.8 Reactor

Coolant Pump Flywheel Inspection Program This program shall provide for inspection of each reactor coolant pump flywheel.At approximately three-year intervals, the bore and keyway of each reactor coolant pump flywheel shall be subjected to an inplace, volumetric examination.

Whenever maintenance or repair activities necessitate flywheel removal, a J "s" u ace xamination of exposed surfaces and a complete volumetric examination

ýctly shall be performed if the interval measured from the previous such inspection is ore greater than 6 2/3 years. The interval may be extended up to one year to permit'ace" inspections to coincide with a planned outage.5.5.9 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves: a. Testing frequencies pecoifg-d i 4 S-cti-n X! of th-1 4AS.E Boier O.d Pa. '.'4c0l CGodc and applicable Addenda as follows: Replace crossed-out text with: "applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code)" OCONEE UNITS 1, 2, & 3 5.0-12 Amendment Nos. 3l, 34., & 340]1 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Inservice Testing Program (continued)

ASME 88ilr HnId PrOUrc Veeeel. Code and applicable Addenda terminology for inservice testing activities Replace crossed-out text with: "1OM"1 Required Frequencies for performing inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provision of SR 3.0.3 are applicable to inservice testing activities; and d. Nothing in the ME PS 44880'Code shall be constru to supersede th requirem ally Replace crossed-out text with: Steam Generator (SG Pr ram FM I ied 5.5.10 V1\ I A Steam Generator Progra shall be established and implemented to ensure that SG tube integrity is mai tained. In addition, the Steam Generator Program shall include the following p visions: a. Provisions for condition m nitoring assessments.

Condition monitoring assessment means an ev auation of the "as found" condition of the tubing with respect to the perform nce criteria for structural integrity and accident induced leakage. The "as f und" condition refers to the condition of the tubing during an SG inspecti n outage, as determined from the inservice inspection results or by other eans, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or lugged to confirm that the performance criteria are being met.Insert after "Frequencies": "and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program" I///OCONEE UNITS 1, 2, & 3 5.0-13 Amendment Nos. [3", 3w, & 94& 11 License Amendment Request No. 2013-04 ATTACHMENT 2 Markups of Technical Specification Bases Pages[14 pages following this cover page]NOTE: Attached are markups of below existing TS Bases pages which incorporate the changes described in the Letter Enclosure.

B 3.4.10-2 and -4 B 3.4.14-6 B 3.5.2-12 and -14 B 3.5.3-8 and -9 B 3.6.5-9 and -11 B 3.7.1-3 and -4 B 3.7.3-4 B 3.7.5-6 and -8 Pressurizer Safety Valves B 3.4.10 BASES (continued)

APPLICABLE SAFETY ANALYSES All accident analyses in the UFSAR that require safety valve actuation assume operation of both pressurizer safety valves to limit increasing reactor coolant pressure.

The overpressure protection analysis is also based on operation of both safety valves and assumes that the valves open at the high range of the setting (2500 psig system design pressure plus 3%). These valves must accommodate pressurizer insurges that could occur during a startup, rod withdrawal, ejected rod, or loss of main feedwater.

The startup accident establishes the minimum safety valve capacity.

The startup accident is assumed to occur at < 15% power.Single failure of a safety valve is neither assumed in the accident analysis nor required to be addressed by the ASME Code. Compliance with this Specification is required to ensure that the accident analysis and design basis calculations remain valid.Pressurizer safety valves satisfy Criterion 3 of 10 CFR 50.36 (Ref. 3).LCO The two pressurizer safety valves are set to open at the RCS design pressure (2500 psig) and within the ASME specified tolerance to avoid exceeding the maximum RCS design pressure SL, to maintain accident analysis assumptions and to comply with ASME Code requirements.

The valves will be tested per ASME -requirements and retumed to service with as-left setpoints of 2500 ps +/- 1%. The upper and lower pressure tolerance limits are based on t requirements of the ASME Boiler and Pressure Vessel Code, Sectio Illl, Article 9, Summer 1967, which limit the rise in pressure within the v ssel which they protect, to 10%above the design pressure.

Inoperability of ne or both valves could result in exceeding the SL if a transient were to oc ur.\The consequences of exceeding the ASME pr ssure limit could include damage to one or more RCS components, incr ased leakage, or additional stress analysis being required prior to of reactor operation.

d n APPLICABILITY In MODES 1, 2, and portions of MODE 3 above th LTOP cut in temperature, OPERABILITY of two valves is require because the combined capacity is required to keep reactor coolan pressure below 110% of its design value during certain accidents.

Po ions of MODE 3 are conservatively included, although the listed accidents ay not require both safety valves for protection.

OCONEE UNITS 1, 2, & 3 B 3.4.10-OCONEE UNIS 1, 2, & B 3.4.10-.&-ex/xx IQ&6WXX/XX/XX 11 m I Pressurizer Safety Valves B 3.4.10 BASES (continued)

SURVEILLANCE REQUIREMENTS SR 3.4.10.1 SRs are specified in the Inservice Testing Program. Pressurizer safety valves are to be tested in accordance with the requirements iem XP of the ASME Code (Ref. 2), which provides the activities an the Frequency necessary to satisfy the SRs. No additional requirements are specified.

The pressurizer safety valves setpoint is +/- 3% for OPERABILITY; however, ['(I the valves are reset to +/-1% during the Surveillance to allow for drift. These values include instrument uncertainties.

f REFERENCES

1. ASME, Boiler and Pressure Vessel Code, Section II1.2. mid P ..... r, VM 0. se .Xh.3. 10 CFR 50.36.Replace Ref. 2 description with: "ASME Code for Operation and Maintenance of Nuclear Power Plants." OCONEE UNITS 1, 2, & 3 3.4.110-4A i4emelmemt C,4. 889339, 333, 3033XX/XX/XX 1

RCS PIV Leakage B 3.4.14 BASES REFERENCES (continued) 7.Aýh Dc*lr -F And PF888kurz 488001o C8@18, 688tWx X6I.Replace Ref. 7 description with: "ASME Code for Operation and Maintenance of Nuclear Power Plants." OCONEE UNITS 1, 2, & 3 B 3.4.14-6-6 ASE REPI.SVIIO.N D,"ATE'D ,O./1.6/+,-

HPI B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verifying the correct alignment for manual and non-automatic power operated valves in the HPI flow paths provides assurance that the proper flow paths will exist for HPI operation.

This SR does apply to the HPI suction header cross-connect valves, the HPI discharge cross-connect valves, the HPI discharge crossover valves, and the LPI-HPI flow path discharge valves (LP-15 and LP-16). This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing.

Similarly, this SR does not apply to automatic valves since automatic valves actuate to their required position upon an accident signal.This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.

The Surveillance Frequency is based on operatinc>-

experience, equipment reliability, and plant risk and is controlled under [I the Surveillance Frequency Control Program.SR 3.5.2.2 With the exception of the HPI pump operating to provide normal makeup, the other two HPI pumps are normally in a standby, non-operating mode.As such, the emergency injection flow path piping has the potential to develop voids and pockets of entrained gases. Venting the HPI pump casings periodically reduces the potential that such voids and pockets of entrained gases can adversely affect operation of the HPI System. This will also reduce the potential for water hammer, pump cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following an ESPS signal. This Surveillance is modified by a Note that indicates it is not applicable to operating HPI pump(s) providing normal makeup. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.SR 3.5.2.3 Periodic surveillance testing of HPI pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by S9t.ieR" a.' the ASME Code (Ref. 5). SRs are specified in the Inservice Testing Program, w'hi-h -no. s Ct., of the ASME Code. ___OCONEE UNITS 1, 2, & 3 B 3.5.2-12 ,A4SES .REI.' .I r.".TED /6/', 3I HPI B 3.5.2 BASES REFERENCES

1. 10 CFR 50.46. Replace Ref. 5 description with: "ASME Code for Operation and 2. UFSAR, Section 15.14.3.3.6.

Maintenance of Nuclear Power Plants."111 I I U .; V. ./ L~~4. NRC Memorandum to V. Stello, Jr., from R.L. aer,"Recommended Interim Revisions to LCOs f r ECCS Components," December 1, 1975.5.A r -ME, B a..... ,,'m ' 3400. ........ -k~p oi ....m- ,....-- 1 l060 IV 6. Letter from R. W. Reid (NRC) to W. 0. Parker, Jr. (Duke)transmitting Safety Evaluation for Oconee Nuclear Station, Units Nos. 1, 2, and 3, Modifications to the High Pressure Injection System, dated December 13, 1978.7. Letter from W. R. McCollum (Duke) to the U. S. NRC, "Proposed Amendment to the Facility Operating License Regarding the High Pressure Injection System Requirements," dated December 16, 1998.OCONEE UNITS 1, 2, & 3 B 3.5.2-14 [BASEG flEVICION BA+ G6/Wi62 1 I LPI B 3.5.3 BASES SURVEILLANCE SR 3.5.3.2 (continued)

REQUIREMENTS cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following an ESPS signal or during shutdown cooling. This Surveillance is modified by a Note that indicates it is not applicable to operating LPI pump(s). The Surveillance Frequency is HT based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. L/J SR 3.5.3.3 Periodic surveillance iesting of LPI pumps to detect gross degradation caused by impeller sliructural damage or other hydraulic component problems is required )y Sooon-X--ehe ASME Code (Ref. 6). SRs are specified in the Inser ice Testing Program. whi'ch Czza", X'I of the ASME Code. \.SR 3.5.3.4 and SR 3.5.3.5 These SRs demonstrate that each automatic LPI valve actuates to the required position on an actual or simulated ESPS signal and that each LPI pump starts on receipt of an actual or simulated ESPS signal. This SR is not required for valves that are locked, sealed, or otherwise secured in position under administrative controls.

The test will be considered satisfactory if control board indication verifies that all components have responded to the ESPS actuation signal properly (all appropriate ESPS actuated pump breakers have opened or closed and all ESPS actuated valves have completed their travel). The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.The actuation logic is tested as part of the ESPS testing, and equipment performance is monitored as part of the Inservice Testing Program.OCONEE UNITS 1, 2, & 3 B 3.5.3-8 BAGEG REYI 0-611 &1 LPI B 3.5.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.3.6 Periodic inspections of the reactor building sump suction inlet ensure that it is unrestricted and stays in proper operating condition.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.REFERENCES

1. 10 CFR 50.46.Replace Ref. 6 description with: "ASME Code for Operation and 2. UFSAR, Section 15.14.3.3.6.

Maintenance of Nu 3. UFSAR, Section 15.14.3.3.5.

Plants." 4. 10 CFR 50.36.5. NRC Memorandum to V. Stello, Jr., from R.L aer,"Recommended Interim Revisions to LCOs r ECCS Components," December 1, 1975.clear Power 6.A&",l Boilo- --A R46664 Vero 'Aifiel Xo6 otoI, Jnoric 7. NRC Safety Evaluation of Babcock & Wilcox Owners Group (B&WOG) Topical Report BAW-2295, Revision 1, "Justification for the Extension of Allowed Outage Time for Low Pressure Injection and Reactor Building Spray systems," (TAC No. MA3807) dated June 30, 1999.OCONEE UNITS 1, 2, & 3 B 3.5.3-9 C D;AES R'-v"SION i

Reactor Building Spray and Cooling Systems B 3.6.5 BASES SURVEILLANCE SR 3.6.5.2 (continued)

REQUIREMENTS Operating each required reactor building cooling train fan unit for_ 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly.

It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.SR 3.6.5.3 Verifying that each required Reactor Building Spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required (iby Code (Ref. 4). Since the Reactor Building Spray Sys em pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance.

Such inservice tests confirm component OPERABILITY, trend performance, and may detect incipient failures by indicating abnormal performance.

The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.6.5.4 Verifying the containment heat removal capability provides assurance that the containment heat removal systems are capable of maintaining containment temperature below design limits following an accident.

This test verifies the heat removal capability of the Low Pressure Injection (LPI)Coolers and Reactor Building Cooling Units. The Surveillance Frequency I is based on operating experience, equipment reliability, and plant risk andII is controlled under the Surveillance Frequency Control Program.OCONEE UNITS 1, 2, & 3 B 3.6.5-9 [jP-PE .EV. 'O XF .T.ED (1/'*1" I]

Reactor Building Spray and Cooling Systems B 3.6.5 BASES REFERENCES 1.2.3.4.UFSAR, Section 3.1.UFSAR, Section 6.2.10 CFR 50.36.Replace crossed-out text with: "ASME Code for Operation and Maintenance of Nuclear Power Plants." e.!occsi Codo_', ipostion Xl.L{pp J OCONEE UNITS 1, 2, & 3 B 3.6.5-11 C 3 .5AiEi RE9'A'RION DTEDW/14/1., II MSRVs B 3.7.1 BASES (continued)

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This SR verifies the OPERABILITY of the MSRVs by the verification of MSRV lift setpoints in accordance with the Inservice Testing Pro ram. The safety and relief valve tests are performed in accordance withANSI Code (Ref. 6) and include the following for MSRVs: a. Visual examination; Change to: b. Seat tightness determination; "ASME" c. Setpoint pressure determination (lift setting);Chan-ae to: "ASME Code" d. Compliance with owner's seat tightness criteria; and e Verification of the balancing device integrity on balanced valves.Thee equires the testing of all valves every 5 years, with a minimum o 20 of the valves tested every 24 months.This SR is modified by a Note that states the surveillance is only required to be performed in MODES 1 and 2. This note allows entry into and operation in MODE 3 prior to performing the SR, provided there is no evidence that the equipment is otherwise believed to be incapable of performing its function.

Also, the guidance in the TS Bases for SR 3.0.1 states that equipment may be considered OPERABLE following maintenance provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function.

This allows operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.

For example, the mode change provisions described above specifically applies to scenarios where maintenance on MSRVs is performed below the mode of applicability for LCO 3.7.1, testing has been satisfactorily completed to the extent possible, and the equipment is believed capable of performing its function.

The mode change provisions permit entry into Mode 3 in order to test and adjust the set pressure, as necessary, to satisfy SR 3.7.1.1 prior to entry into Mode 2.The MSRVs may be either bench tested or tested in situ at hot conditions using an assist device to simulate lift pressure.

If the MSRVs are not tested at hot conditions, the lift setting pressure must be corrected to ambient conditions of the valve at operating temperature and pressure.OCONEE UNITS 1, 2, & 3 B 3.7.1-3 DAts ^ns..... D"T"D I ]

MSRVs B 3.7.1 BASES (continued)

REFERENCES 1.2.3.4.5.6.UFSAR, Section 10.3.ASME, Boiler and Pressure Vessel Code,Section III, Article NC-7000, Class 2 Components.

UFSAR, Chapter 15. Replace crossed-out text with: "ASME Code for Operation and UFSAR, Section 10.3.3. Maintenance of Nuclear Power Plants." 10 CFR 50.36.AWI.AA~ S A .19-- PrzzFurz 8. 1 Gods, roctin 'XI.I OCONEE UNITS 1, 2, & 3 B 3.7.1-4 I AGEG RlEVISION D)ATED $ 061. ij MFCVs and SFCVs B 3.7.3 BASES (continued)

SURVEILLANCE REQUIREMENTS SR 3.7.3.1 This SR verifies that the closure time of each MFCV and SFCV is_< 25 seconds on an actual or simulated actuation signal. The 25 seconds includes a 10 second signal delay and 15 seconds for valve movement.The MFCV and SFCV closure time is assumed in the containment analyses.

This Surveillance is normally performed upon returning the unit to operation following a refueling outage. The MFCV and SFCV should not be tested at power since even a part stroke exercise increases the risk of a valve closure with the unit generating power. This is consistent with the ASME Coe (t7, Ref. 2) requirements during operation in MODES 1 and 2.This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR.The Frequency for this SR is in accordance with the Inservice Testing Program.REFERENCES

1. 10 CFR 50.36.2. ACG..E, 9851OF B.-z rrczzUrz'FoS Cdo oto I Replace crossed-out text with: "ASME Code for Operation and Maintenance of Nuclear Power Plants." OCONEE UNITS 1, 2, & 3 B 3.7.3-4[ DACEf~ flEVIC~ON DATED ~l~I9~ I EFW System B 3.7.5 BASES (continued)

SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, and non-automatic power operated valves in the EFW water and steam supply flow paths provides assurance that the proper flow paths exist for EFW operation.

This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since those valves are verified to be in the correct position prior to locking, sealing, or securing.This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.SR 3.7.5.2 Verifying that each EFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that EFW pump performance has not degraded below the acceptance criteria during the cycle. Flow and dilerential head are ormal indications of pump performance required y &e_%@.Xl,4vthe ASME Code (Ref. 3). Because it is undesirable to intr uce cold EFW inmt the steam generators while they are operating, this test may be performed on a test flow path.This test confirms OPERABILITY, trends performance, and detects incipient failures by indicating abnormal nerformance.

Performance of inservice testing in the ASME ode,-&@eet.§. .V(Ref. 3), at 3 month intervals, satisfies this requirement.

SR 3.7.5.3 This SR verifies that EFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an Emergency Feedwater System initiation signal by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal. This SR is not required for valves that are locked, sealed, or otherwise secured in position under administrative OCONEE UNITS 1, 2, & 3 B 3.7.5-6 [jAe1 fE./eO A[ 1.1 i EFW System B 3.7.5 BASES (continued)

REFERENCES

1. UFSAR, Section 10.4.7.2. 10 CFR 50.36.3. AQAA, _p9Igr nd P;-occ:- \-'ccol Cod3, -4ztizrm Al.Replace crossed-out text with: "ASME Code for Operation and Maintenance of Nuclear Power Plants." OCONEE UNITS 1, 2, & 3 B 3.7.5-8 9 U3A REVS EV',',O, DPATED GS61.I4 License Amendment Request No. 2013-04 ATTACHMENT 3 Revised Technical Specification Pages[2 pages following this cover page]NOTE: Attached are clean, retyped TS Pages 5.0-12 and -13.

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.7 Pre-Stressed Concrete Containment Tendon Surveillance Pro-gram This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity.

The program shall include baseline measurements prior to initial operations.

The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Section Xl, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, as amended by relief granted in accordance with 10 CFR 50.55a(a)(3).

The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

5.5.8 Reactor

Coolant Pump Flywheel Inspection Program This program shall provide for inspection of each reactor coolant pump flywheel.At approximately three-year intervals, the bore and keyway of each reactor coolant pump flywheel shall be subjected to an inplace, volumetric examination.

Whenever maintenance or repair activities necessitate flywheel removal, a surface examination of exposed surfaces and a complete volumetric examination shall be performed if the interval measured from the previous such inspection is greater than 6 2/3 years. The interval may be extended up to one year to permit inspections to coincide with a planned outage.5.5.9 Inservice Testinq Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves: a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows: OCONEE UNITS 1, 2, & 3 5.0-12 Amendment Nos. _, _, & __ I Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Inservice Testing Program (continued)

ASME OM Code and applicable Addenda terminology for inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days 5.5.10 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 Inservice 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.Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained.

In addition, the Steam Generator Program shall include the following provisions:

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.OCONEE UNITS 1, 2, & 3 5.0-13 Amendment Nos. -, -, & I License Amendment Request No. 2013-04 ATTACHMENT 4 Revised Technical Specification Bases Pages (14 pages following this cover page]NOTE: Attached are clean, retyped TS Bases pages, as specified below.B 3.4.10-2 and -4 B 3.4.14-6 B 3.5.2-12 and -14 B 3.5.3-8 and -9 B 3.6.5-9 and -11 B 3.7.1-3 and -4 B 3.7.3-4 B 3.7.5-6 and -8 Pressurizer Safety Valves B 3.4.10 BASES (continued)

APPLICABLE SAFETY ANALYSES All accident analyses in the UFSAR that require safety valve actuation assume operation of both pressurizer safety valves to limit increasing reactor coolant pressure.

The overpressure protection analysis is also based on operation of both safety valves and assumes that the valves open at the high range of the setting (2500 psig system design pressure plus 3%). These valves must accommodate pressurizer insurges that could occur during a startup, rod withdrawal, ejected rod, or loss of main feedwater.

The startup accident establishes the minimum safety valve capacity.

The startup accident is assumed to occur at < 15% power.Single failure of a safety valve is neither assumed in the accident analysis nor required to be addressed by the ASME Code. Compliance with this Specification is required to ensure that the accident analysis and design basis calculations remain valid.Pressurizer safety valves satisfy Criterion 3 of 10 CFR 50.36 (Ref. 3).LCO The two pressurizer safety valves are set to open at the RCS design pressure (2500 psig) and within the ASME specified tolerance to avoid exceeding the maximum RCS design pressure SL, to maintain accident analysis assumptions and to comply with ASME Code requirements.

The valves will be tested per ASME Code requirements and returned to service with as-left setpoints of 2500 psig +/- 1%. The upper and lower pressure tolerance limits are based on the requirements of the ASME Boiler and Pressure Vessel Code,Section III, Article 9, Summer 1967, which limit the rise in pressure within the vessel which they protect, to 10% above the design pressure.

Inoperability of one or both valves could result in exceeding the SL if a transient were to occur.The consequences of exceeding the ASME pressure limit could include damage to one or more RCS components, increased leakage, or additional stress analysis being required prior to resumption of reactor operation.

APPLICABILITY In MODES 1, 2, and portions of MODE 3 above the LTOP cut in temperature, OPERABILITY of two valves is required because the combined capacity is required to keep reactor coolant pressure below 110% of its design value during certain accidents.

Portions of MODE 3 are conservatively included, although the listed accidents may not require both safety valves for protection.

OCONEE UNITS 1, 2, & 3 B 3.4.10-2 xx/xx/xx I Pressurizer Safety Valves B 3.4.10 BASES (continued)

SURVEILLANCE REQUIREMENTS SR 3.4.10.1 SRs are specified in the Inservice Testing Program. Pressurizer safety valves are to be tested in accordance with the requirements of the ASME Code (Ref. 2), which provides the activities and the Frequency necessary to satisfy the SRs. No additional requirements are specified.

The pressurizer safety valves setpoint is +/- 3% for OPERABILITY; however, the valves are reset to +/-1% during the Surveillance to allow for drift. These values include instrument uncertainties.

REFERENCES

1. ASME, Boiler and Pressure Vessel Code,Section III.2. ASME Code for Operation and Maintenance of Nuclear Power Plants.3. 10 CFR 50.36.OCONEE UNITS 1, 2, & 3 B 3.4.10-4 XX/XX/XX I RCS PIV Leakage B 3.4.14 BASES REFERENCES

7. ASME Code for Operation and Maintenance of Nuclear Power Plants.(continued)

OCONEE UNITS 1, 2, & 3 B 3.4.14-6 XX/XX/XX I HPI B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verifying the correct alignment for manual and non-automatic power operated valves in the HPI flow paths provides assurance that the proper flow paths will exist for HPI operation.

This SR does apply to the HPI suction header cross-connect valves, the HPI discharge cross-connect valves, the HPI discharge crossover valves, and the LPI-HPI flow path discharge valves (LP-15 and LP-16). This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing.

Similarly, this SR does not apply to automatic valves since automatic valves actuate to their required position upon an accident signal.This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.SR 3.5.2.2 With the exception of the HPI pump operating to provide normal makeup, the other two HPI pumps are normally in a standby, non-operating mode.As such, the emergency injection flow path piping has the potential to develop voids and pockets of entrained gases. Venting the HPI pump casings periodically reduces the potential that such voids and pockets of entrained gases can adversely affect operation of the HPI System. This will also reduce the potential for water hammer, pump cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following an ESPS signal. This Surveillance is modified by a Note that indicates it is not applicable to operating HPI pump(s) providing normal makeup. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.SR 3.5.2.3 Periodic surveillance testing of HPI pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code (Ref. 5). SRs are specified in the Inservice Testing Program of the ASME Code.OCONEE UNITS 1, 2, & 3 B 3.5.2-12 XX/XX/XX I HPI B 3.5.2 BASES REFERENCES

1. 10 CFR 50.46.2. UFSAR, Section 15.14.3.3.6.

3. 10 CFR 50.36.4. NRC Memorandum to V. Stello, Jr., from R.L. Baer,"Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.5. ASME Code for Operation and Maintenance of Nuclear Power Plants.6. Letter from R. W. Reid (NRC) to W. 0. Parker, Jr. (Duke)transmitting Safety Evaluation for Oconee Nuclear Station, Units Nos. 1, 2, and 3, Modifications to the High Pressure Injection System, dated December 13, 1978.7. Letter from W. R. McCollum (Duke) to the U. S. NRC, "Proposed Amendment to the Facility Operating License Regarding the High Pressure Injection System Requirements," dated December 16, 1998.OCONEE UNITS 1, 2, & 3 B 3.5.2-14 XX/XX/XX I LPI B 3.5.3 BASES SURVEILLANCE SR 3.5.3.2 (continued)

REQUIREMENTS cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following an ESPS signal or during shutdown cooling. This Surveillance is modified by a Note that indicates it is not applicable to operating LPI pump(s). The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.SR 3.5.3.3 Periodic surveillance testing of LPI pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code (Ref. 6). SRs are specified in the Inservice Testing Program of the ASME Code.SR 3.5.3.4 and SR 3.5.3.5 These SRs demonstrate that each automatic LPI valve actuates to the required position on an actual or simulated ESPS signal and that each LPI pump starts on receipt of an actual or simulated ESPS signal. This SR is not required for valves that are locked, sealed, or otherwise secured in position under administrative controls.

The test will be considered satisfactory if control board indication verifies that all components have responded to the ESPS actuation signal properly (all appropriate ESPS actuated pump breakers have opened or closed and all ESPS actuated valves have completed their travel). The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.The actuation logic is tested as part of the ESPS testing, and equipment performance is monitored as part of the Inservice Testing Program.OCONEE UNITS 1, 2, & 3 B 3.5.3-8 XX/XX/XX I LPI B 3.5.3 BASES SURVEILLANCE SR 3.5.3.6 REQUIREMENTS (continued)

Periodic inspections of the reactor building sump suction inlet ensure that it is unrestricted and stays in proper operating condition.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.REFERENCES

1. 10 CFR 50.46.2. UFSAR, Section 15.14.3.3.6.

3. UFSAR, Section 15.14.3.3.5.

4. 10 CFR 50.36.5. NRC Memorandum to V. Stello, Jr., from R.L. Baer,"Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.6. ASME Code for Operation and Maintenance of Nuclear Power Plants.7. NRC Safety Evaluation of Babcock & Wilcox Owners Group (B&WOG) Topical Report BAW-2295, Revision 1, "Justification for the Extension of Allowed Outage Time for Low Pressure Injection and Reactor Building Spray systems," (TAC No. MA3807) dated June 30, 1999.OCONEE UNITS 1, 2, & 3 B 3.5.3-9 XX/XX/XX I Reactor Building Spray and Cooling Systems B 3.6.5 BASES SURVEILLANCE SR 3.6.5.2 (continued)

REQUIREMENTS Operating each required reactor building cooling train fan unit for>_ 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly.

It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.SR 3.6.5.3 Verifying that each required Reactor Building Spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by the ASME Code (Ref. 4). Since the Reactor Building Spray System pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance.

Such inservice tests confirm component OPERABILITY, trend performance, and may detect incipient failures by indicating abnormal performance.

The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.6.5.4 Verifying the containment heat removal capability provides assurance that the containment heat removal systems are capable of maintaining containment temperature below design limits following an accident.

This test verifies the heat removal capability of the Low Pressure Injection (LPI)Coolers and Reactor Building Cooling Units. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.OCONEE UNITS 1, 2, & 3 B 3.6.5-9 XX/XX/XX I Reactor Building Spray and Cooling Systems B 3.6.5 BASES REFERENCES

1. UFSAR, Section 3.1.2. UFSAR, Section 6.2.3. 10 CFR 50.36.4. ASME Code for Operation and Maintenance of Nuclear Power Plants.OCONEE UNITS 1, 2, & 3 B 3.6.5-11 XX/XX/XX I MSRVs B 3.7.1 BASES (continued)

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This SR verifies the OPERABILITY of the MSRVs by the verification of MSRV lift setpoints in accordance with the Inservice Testing Program. The safety and relief valve tests are performed in accordance with ASME Code (Ref. 6) and include the following for MSRVs: a. Visual examination;

b. Seat tightness determination;

c. Setpoint pressure determination (lift setting);d. Compliance with owner's seat tightness criteria; and e. Verification of the balancing device integrity on balanced valves.The ASME Code requires the testing of all valves every 5 years, with a minimum of 20% of the valves tested every 24 months.This SR is modified by a Note that states the surveillance is only required to be performed in MODES I and 2. This note allows entry into and operation in MODE 3 prior to performing the SR, provided there is no evidence that the equipment is otherwise believed to be incapable of performing its function.

Also, the guidance in the TS Bases for SR 3.0.1 states that equipment may be considered OPERABLE following maintenance provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function.

This allows operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.

For example, the mode change provisions described above specifically applies to scenarios where maintenance on MSRVs is performed below the mode of applicability for LCO 3.7.1, testing has been satisfactorily completed to the extent possible, and the equipment is believed capable of performing its function.

The mode change provisions permit entry into Mode 3 in order to test and adjust the set pressure, as necessary, to satisfy SR 3.7.1.1 prior to entry into Mode 2.The MSRVs may be either bench tested or tested in situ at hot conditions using an assist device to simulate lift pressure.

If the MSRVs are not tested at hot conditions, the lift setting pressure must be corrected to ambient conditions of the valve at operating temperature and pressure.OCONEE UNITS 1, 2, & 3 B 3.7.1-3 XX/XX/XX I MSRVs B 3.7.1 BASES (continued)

REFERENCES 1.2.3.4.5.6.UFSAR, Section 10.3.ASME, Boiler and Pressure Vessel Code,Section III, Article NC-7000, Class 2 Components.

UFSAR, Chapter 15.UFSAR, Section 10.3.3.10 CFR 50.36.ASME Code for Operation and Maintenance of Nuclear Power Plants.OCONEE UNITS 1, 2, & 3 B 3.7.1-4 XX/XX/XX I MFCVs and SFCVs B 3.7.3 BASES (continued)

SURVEILLANCE REQUIREMENTS SR 3.7.3.1 This SR verifies that the closure time of each MFCV and SFCV is_< 25 seconds on an actual or simulated actuation signal. The 25 seconds includes a 10 second signal delay and 15 seconds for valve movement.The MFCV and SFCV closure time is assumed in the containment analyses.

This Surveillance is normally performed upon returning the unit to operation following a refueling outage. The MFCV and SFCV should not be tested at power since even a part stroke exercise increases the risk of a valve closure with the unit generating power. This is consistent with the ASME Code (Ref. 2) requirements during operation in MODES 1 and 2.This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR.The Frequency for this SR is in accordance with the Inservice Testing Program.REFERENCES

1. 10 CFR 50.36.2. ASME Code for Operation and Maintenance of Nuclear Power Plants.OCONEE UNITS 1, 2, & 3 B 3.7.3-4 XX/XX/XX I EFW System B 3.7.5 BASES (continued)

SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, and non-automatic power operated valves in the EFW water and steam supply flow paths provides assurance that the proper flow paths exist for EFW operation.

This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since those valves are verified to be in the correct position prior to locking, sealing, or securing.This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.SR 3.7.5.2 Verifying that each EFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that EFW pump performance has not degraded below the acceptance criteria during the cycle. Flow and differential head are normal indications of pump performance required by the ASME Code (Ref. 3). Because it is undesirable to introduce cold EFW into the steam generators while they are operating, this test may be performed on a test flow path.This test confirms OPERABILITY, trends performance, and detects incipient failures by indicating abnormal performance.

Performance of inservice testing in the ASME Code (Ref. 3), at 3 month intervals, satisfies this requirement.

SR 3.7.5.3 This SR verifies that EFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an Emergency Feedwater System initiation signal by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal. This SR is not required for valves that are locked, sealed, or otherwise secured in position under administrative OCONEE UNITS 1, 2, & 3 B 3.7.5-6 XX/XXIXX IlOCONEE UNITS 1, 2, & 3 B 3.7.5-6 XX/XX/XX I EFW System B 3.7.5 BASES (continued)

REFERENCES

1. UFSAR, Section 10.4.7.2. 10 CFR 50.36.3. ASME Code for Operation and Maintenance of Nuclear Power Plants.OCONEE UNITS 1, 2, & 3 B 3.7.5-8 XX/XX/XX I