ML14078A037

Oconee, Units 1, 2 and 3 - 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. BatonVice President ENERGY, Oconee Nuclear StationDuke EnergyONO1VP 1 7800 Rochester HwySeneca, SC 29672ONS-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 DeskU. S. Nuclear Regulatory Commission 11555 Rockville PikeRockville, MD 20852Duke Energy Carolinas, LLC (Duke Energy)Oconee Nuclear Station, Units 1, 2, and 3Docket Nos. 50-269, 50-270, and 50-287Renewed License Nos. DPR-38, DPR-47, and DPR-55

Subject:

License Amendment Request (LAR) for Adoption of Technical Specification TaskForce (TSTF) Change Travelers TSTF-479 and TSTF-497Oconee Nuclear Station (ONS) LAR No. 2013-04In accordance with the provisions of Section 50.90 of Title 10 of the Code of FederalRegulations (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 proposedamendment 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 Societyof Mechanical Engineers (ASME) Code that is referenced in 10 CFR 50.55a(b).

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

The adoption of the latest editionrenders incorrect certain statements in the TS Administrative Controls Inservice TestingProgram and in the TS Bases. These changes are consistent with NRC-approved Revision 0 toTSTF Improved Standard Technical Specification Change Travelers TSTF-479, "Changes toReflect Revision of 10 CFR 50.55a,"

and Revision 0 to TSTF-497, "Limit Inservice TestingProgram 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 providemark-ups of the corrected TS and TS Bases pages, respectively.

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

Once this amendment request is approved, the amendment will be implemented within 120days. Duke Energy will also update applicable sections of the ONS Updated Final SafetyAnalysis Report (UFSAR),

as necessary, and submit the updated UFSAR sections inaccordance with 10 CFR 50.71(e).

There are no new regulatory commitments being made as aresult of the proposed change.X06c1 U. S. Nuclear Regulatory Commission March 14, 2014Page 2If there are any questions regarding the content of this document or if additional information isneeded, 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 14thday of March, 2014.Sincerely, Scott L. BatsonSite Vice President Oconee Nuclear Station

Enclosure:

Evaluation of the Proposed ChangesAttachments:

1. Attachment 1 -Markups of Technical Specification Pages2. Attachment 2 -Markups of Technical Specification Bases Pages3. Attachment 3 -Revised Technical Specification Pages4. Attachment 4 -Revised Technical Specification Bases Pages U. S. Nuclear Regulatory Commission March 14, 2014Page 3cc w/enclosure and attachments:

Mr. Victor McCreeAdministrator, Region IIU.S. Nuclear Regulatory Commission Marquis One Tower245 Peachtree Center Ave., NE, Suite 1200Atlanta, GA 30303-1257 Mr. Richard GuzmanSenior Project Manager(by electronic mail only)U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation 11555 Rockville PikeMail Stop O-8C2Rockville, MD 20852Mr. Eddy CroweNRC Senior Resident Inspector Oconee Nuclear StationMs. Susan E. Jenkins,

Manager, Infectious and Radioactive Waste Management, Division of Waste Management South Carolina Department of Health & Environmental Control2600 Bull Street,Columbia, SC 29201 ENCLOSURE EVALUATION OF THE PROPOSED CHANGESLICENSE AMENDMENT REQUEST NO. 2013-04

Subject:

License Amendment Request for the Adoption of Technical Specification TaskForce (TSTF) Change Travelers TSTF-479 and TSTF-4971 SUMMARY DESCRIPTION 2 BACKGROUND 3 DESCRIPTION OF PROPOSED CHANGES4 TECHNICAL ANALYSIS5 REGULATORY ANALYSIS5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

5.3 Precedence

6 ENVIRONMENTAL CONSIDERATION 7 REFERENCES Enclosure

-Evaluation of Proposed ChangesLicense Amendment Request No. 2013-04 Page 1 of 91 SUMMARY DESCRIPTION The proposed amendment would revise the Oconee Nuclear Station (ONS) Units 1, 2, and 3Technical Specifications (TS) Administrative Controls Inservice Testing Program (i.e., TS5.5.9) and references in the TS Bases (TSB) to reflect the current edition of the AmericanSociety of Mechanical Engineers (ASME) Code that is referenced in 10 CFR 50.55a(b).

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

Theadoption of the latest edition renders select statements in the TS Administrative ControlsITP 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) andTSTF-497, "Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Yearsor 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 TSBases pages associated with this license amendment request (LAR) are provided inAttachments 1 and 2, respectively, and the retyped (clean) TS and TSB pages are providedin Attachments 3 and 4, respectively.

Once this LAR is approved, the amendment will be implemented within 120 days. There areno 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 powerplants. The ASME intended that the ASME OM Code replace Section XI of the ASME Boilerand Pressure Vessel (B&PV) Code for inservice testing of pumps and valves. The 1995edition of the ASME OM Code was incorporated by reference into 10 CFR 50.55a onSeptember 22, 1999 (Ref. 3). Since 10 CFR 50.55a(f)(4)(ii) requires that inservice testingcomply with the requirements of the latest edition and addenda of the ASME Codeincorporated into 10 CFR 50.55a(b),

TS 5.5.9 must be revised to reference the ASME OMCode. TSTF-479 was developed to provide licensees a standard method to request NRCapproval 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 datedDecember 6, 2005 (Ref. 4).Although the NRC approved TSTF-479, the NRC expressed concerns with the TSTFchanges 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 applyingSurveillance Requirement (SR) 3.0.2 to Frequencies specified in the ITP as greater than twoyears. Thus, the TSTF Group developed TSTF-497 to revise paragraph b of the ITP TS to Enclosure

-Evaluation of Proposed ChangesLicense Amendment Request No. 2013-04 Page 2 of 9specify 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 CHANGESDuke Energy proposes to modify the TS and TSB (for information only). The proposedchange to ONS TS 5.5.8 only corrects a typographical error. The proposed changes toONS TS 5.5.9, and in the TSB, will replace reference to ASME B&PV Code with reference toASME OM Code for pump and valve testing only. The TS 5.5.9 proposed changes adoptchanges specified in NRC-approved TSTF-479 and TSTF-497 without variations ordeviations.

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 VesselCode" with "applicable to the ASME Code for Operation and Maintenance of NuclearPower Plants (ASME OM Code)."" TS 5.5.9a -In the column heading that states "ASME Boiler and Pressure VesselCode 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:

"andto other normal and accelerated Frequencies specified as 2 years or less in theInservice 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 XIrequirements..

." to ".... per ASME Code requirements..."

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

Enclosure

-Evaluation of Proposed ChangesLicense Amendment Request No. 2013-04 Page 3 of 9For 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 wordingto state" "SRs are specified in the Inservice Testing Program of the ASME Code."* References Section -Change Reference 5 to state "ASME Code for Operation andMaintenance 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 wordingto state" "SRs are specified in the Inservice Testing Program of the ASME Code."" References Section -Change Reference 6 to state "ASME Code for Operation andMaintenance 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 andMaintenance 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 andMaintenance 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 andMaintenance 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 andMaintenance of Nuclear Power Plants."

Enclosure

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

4 TECHNICAL ANALYSISThe 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 safetymargins with provisions for increased surveillance and corrective action. NRC regulation 10 CFR 50.55a defines the requirements for applying industry codes to each licensednuclear 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 theCode 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 toprovide updated requirements for the inservice inspection and inservice testing ofcomponents.

Until 1990, the ASME Code requirements addressing the inservice testing ofpumps and valves were contained in Section Xl, Subsections IWP (for pumps) and IWV (forvalves).

In 1990, the ASME published the initial edition of the OM Code that provides therules for inservice testing of pumps and valves. Since the establishment of the 1990 Editionof the OM Code, the rules for inservice testing of pumps are no longer being updated inSection XI. As identified in NRC SECY-99-017, dated January 13, 1999 (Ref. 6), the NRChas generally considered the evolution of the ASME Code to result in a net improvement inthe 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 otherInservice Testing Frequencies, of two years or less, that are not specified in the ITP. TheITP may have Frequencies for testing that are based on risk and do not conform to thestandard testing Frequencies specified in the TS. For example, an ITP may use ASMECode Case OMN-1, "Alternative Rules for Preservice and Inservice Testing of CertainElectric Motor-Operated Valve Assemblies in Light-Water Reactor Plants,"

in lieu of stroketime testing.

The Frequency of the Surveillance may be determined through a mix of riskinformed and performance based means in accordance with the ITP. This is consistent withthe 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.

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

Enclosure

-Evaluation of Proposed ChangesLicense Amendment Request No. 2013-04 Page 5 of 95 REGULATORY ANALYSIS5.1 NO SIGNIFICANT HAZARDS CONSIDERATION Duke Energy Carolinas, LLC (Duke Energy),

has evaluated the proposed changes to theOconee Nuclear Station (ONS) Technical Specifications (TS) using the criteria in10 CFR 50.92 and has determined that the proposed changes do not involve asignificant hazards consideration.

An analysis of the issue of no significant hazardsconsideration is presented below:Description of Amendment ReauestThe proposed amendment would correct a typographical error in TS 5.5.8, "ReactorCoolant Pump Flywheel Inspection Program,"

and revise TS 5.5.9, "Inservice TestingProgram,"

to include testing frequencies applicable to the American Society ofMechanical Engineers (ASME) Operation and Maintenance (OM) Code instead of theASME Boiler and Pressure Vessel (B&PV) Code,Section XI. Additionally, TS 5.5.9would 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 thethree 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 orConsequences of an Accident Previously Evaluated?

Response:

NoThe proposed change corrects a typographical error in TS 5.5.8, "ReactorCoolant 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 asASME 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 fortesting pumps and valves.The proposed change does not impact any accident initiators or analyzed eventsor assumed mitigation of accident or transient events. The proposed changedoes not involve the addition or removal of any equipment, or any designchanges to the facility.

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

Enclosure

-Evaluation of Proposed ChangesLicense Amendment Request No. 2013-04 Page 6 of 92. Does the Proposed Change Create the Possibility of a New or Different Kind ofAccident from any Accident Previously Evaluated?

Response:

NoThe proposed change corrects a typographical error in TS 5.5.8, "ReactorCoolant 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 asASME 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 fortesting pumps and valves.The proposed change does not involve a modification to the physicalconfiguration of the plant (i.e., no new equipment will be installed),

nor does itinvolve a change in the methods governing normal plant operation.

Theproposed 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 anyeffluent that may be released offsite and there is no increase in individual orcumulative occupational exposure.

Therefore, the proposed change does notcreate the possibility of a new or different kind of accident from any accidentpreviously evaluated.

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

NoThe proposed change corrects a typographical error in TS 5.5.8, "ReactorCoolant 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 asASME 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 fortesting pumps and valves. The safety function of the affected pumps and valveswill be maintained.

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

"Issuance of Amendment."

Enclosure

-Evaluation of Proposed ChangesLicense Amendment Request No. 2013-04 Page 7 of 95.2 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA NRC regulation 10 CFR 50.55a defines the requirements for applying industry codes toeach licensed nuclear powered facility.

Licensees are required by 10 CFR50.55a(f)(4)(i) to initially prepare programs to perform inservice testing of certain ASMESection 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 editionand addenda incorporated into paragraph (b) of 10 CFR 50.55a on the date 12 monthsprior to the date of issuance of the operating license subject to the limitations andmodification 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 thecontainment.

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

Therefore, basedon the considerations discussed above:1) There is reasonable assurance that the health and safety of the public will not beendangered by operation in the proposed manner;2) Such activities will be conducted in compliance with the Commission's regulations; and3) Issuance of the amendment will not be inimical to the common defense andsecurity 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 nuclearpower plants resulted in the following documents of precedence.

• NRC Letter to Nuclear Management

Company, LLC, "Prairie Island NuclearGenerating Plant, Units 1 and 2 -Issuance of Amendments Re: Incorporation ofTechnical Specification Task Force Travelers TSTF-479, TSTF-485 andTSTF-497 (TAC Nos. MD5983 and MD5984),"

dated June 27, 2008 [ADAMSAccession No. ML081650272].

" NRC Letter to Exelon Generation

Company, LLC, "Braidwood

Station, Units 1and 2; Byron Station, Units Nos. 1 and 2; Dresden Nuclear Power Station, Units2 and 3; Limerick Generating

Station, Units 1 and 2; Oyster Creek NuclearGenerating Station; Peach Bottom Atomic Power Station, Units 2 and 3; QuadCities Nuclear Power Station, Units 1 and 2; and Three Mile Island NuclearStation, 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 ChangesLicense Amendment Request No. 2013-04 Page 8 of 9* NRC Letter to Duke Energy Carolinas, LLC, "McGuire Nuclear Station, Units 1and 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. MD9581and MD9582),"

dated August 17, 2009 [ADAMS Accession No. ML092240085].

• NRC Letter to Duke Energy Carolinas, LLC, "Catawba Nuclear Station, Units 1and 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 ToReflect Revision To 10 CFR 50.55a, Technical Specification Task Force ChangeTravelers TSTF-479-A and TSTF-497-A (TAC No. ME1829),"

dated December23, 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 ToReflect 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, andME3916),"

dated January 19, 2011 [ADAMS Accession No. MLI103560088].

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

However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in theamounts of any effluents that may be released

offsite, or (iii) a significant increase inindividual or cumulative occupational radiation exposure.

Accordingly, the proposed changemeets 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 orenvironmental 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, 'Changesto 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"' [ADAMSAccession No. ML061930221].

Enclosure

-Evaluation of Proposed ChangesLicense Amendment Request No. 2013-04 Page 9 of 93. Federal Register Notice 64 FR 51370, dated September 22, 1999, "Industry Codes andStandards; Amended Requirements."

4. NRC Letter to Technical Specifications Task Force, dated December 6, 2005, "Status ofTSTF 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 CFR50.55a."7. NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants,"

Revision 2(Draft Report for Comment),

August 2011.

License Amendment Request No. 2013-04ATTACHMENT IMarkups 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 thechanges described in the Letter Enclosure.

Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.7 Pre-Stressed Concrete Containment Tendon Surveillance ProgramThis program provides controls for monitoring any tendon degradation inpre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity.

The program shallinclude baseline measurements prior to initial operations.

The TendonSurveillance

Program, inspection frequencies, and acceptance criteria shall be inaccordance with Section XI, Subsection IWL of the ASME Boiler and PressureVessel Code and applicable addenda as required by 10 CFR 50.55a, asamended 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 Programinspection frequencies.

5.5.8 Reactor Coolant Pump Flywheel Inspection ProgramThis program shall provide for inspection of each reactor coolant pump flywheel.

At approximately three-year intervals, the bore and keyway of each reactorcoolant pump flywheel shall be subjected to an inplace, volumetric examination.

Whenever maintenance or repair activities necessitate flywheel

removal, aJ "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 isore 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 ProgramThis 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.dPa. '.'4c0l CGodc and applicable Addenda as follows:Replace crossed-out text with:"applicable to the ASME Code for Operations and Maintenance ofNuclear Power Plants (ASME OM Code)"OCONEE UNITS 1, 2, & 3 5.0-12 Amendment Nos. 3l, 34., & 340]1 Programs and Manuals5.55.5 Programs and Manuals5.5.9Inservice Testing Program (continued)

ASME 88ilr HnId PrOUrcVeeeel. Code andapplicable Addendaterminology forinservice testingactivities Replace crossed-out text with:"1OM"1Required Frequencies for performing inservice testing activities WeeklyMonthlyQuarterly or every3 monthsSemiannually orevery 6 monthsEvery 9 monthsYearly or annuallyBiennially or every 2 yearsAt least once per 7 daysAt least once per 31 daysAt least once per 92 daysAt least once per 184 daysAt least once per 276 daysAt least once per 366 daysAt least once per 731 daysb. The provisions of SR 3.0.2 are applicable to the above requiredFrequencies for performing inservice testing activities;

c. The provision of SR 3.0.3 are applicable to inservice testing activities; andd. Nothing in the ME PS 44880'Code shall be construto supersede th requirem allyReplace crossed-out text with:Steam Generator (SG Pr ram FM Iied5.5.10V1\ IA Steam Generator Progra shall be established and implemented to ensurethat SG tube integrity is mai tained. In addition, the Steam Generator Programshall 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 tubingwith respect to the perform nce criteria for structural integrity and accidentinduced leakage.

The "as f und" condition refers to the condition of thetubing 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 whichthe 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 theInservice Testing Program"I///OCONEE UNITS 1, 2, & 35.0-13Amendment Nos. [3", 3w, & 94& 11 License Amendment Request No. 2013-04ATTACHMENT 2Markups of Technical Specification Bases Pages[14 pages following this cover page]NOTE: Attached are markups of below existing TS Bases pages which incorporate thechanges described in the Letter Enclosure.

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

APPLICABLE SAFETY ANALYSESAll accident analyses in the UFSAR that require safety valveactuation assume operation of both pressurizer safety valves to limitincreasing reactor coolant pressure.

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

The startup accident establishes the minimum safetyvalve capacity.

The startup accident is assumed to occur at < 15% power.Single failure of a safety valve is neither assumed in the accident analysisnor required to be addressed by the ASME Code. Compliance with thisSpecification is required to ensure that the accident analysis and designbasis calculations remain valid.Pressurizer safety valves satisfy Criterion 3 of 10 CFR 50.36 (Ref. 3).LCOThe two pressurizer safety valves are set to open at the RCS designpressure (2500 psig) and within the ASME specified tolerance to avoidexceeding the maximum RCS design pressure SL, to maintain accidentanalysis assumptions and to comply with ASME Code requirements.

Thevalves will be tested per ASME -requirements and retumed toservice with as-left setpoints of 2500 ps +/- 1%. The upper and lowerpressure tolerance limits are based on t requirements of the ASMEBoiler 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 resultin exceeding the SL if a transient were to oc ur.\The consequences of exceeding the ASME pr ssure limit could includedamage to one or more RCS components, incr ased leakage, or additional stress analysis being required prior to of reactor operation.

d nAPPLICABILITY In MODES 1, 2, and portions of MODE 3 above th LTOP cut intemperature, OPERABILITY of two valves is require because thecombined capacity is required to keep reactor coolan pressure below110% of its design value during certain accidents.

Po ions of MODE 3 areconservatively

included, although the listed accidents ay not require bothsafety valves for protection.

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

SURVEILLANCE REQUIREMENTS SR 3.4.10.1SRs are specified in the Inservice Testing Program.

Pressurizer safetyvalves are to be tested in accordance with the requirements iem XPof the ASME Code (Ref. 2), which provides the activities an theFrequency necessary to satisfy the SRs. No additional requirements arespecified.

The pressurizer safety valves setpoint is +/- 3% for OPERABILITY;

however,

['(Ithe valves are reset to +/-1% during the Surveillance to allow for drift. Thesevalues include instrument uncertainties.

fREFERENCES

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 andMaintenance of Nuclear Power Plants."OCONEE UNITS 1, 2, & 33.4.110-4A i4emelmemt C,4. 889339, 333, 3033XX/XX/XX 1

RCS PIV LeakageB 3.4.14BASESREFERENCES (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 NuclearPower Plants."OCONEE UNITS 1, 2, & 3B 3.4.14-6-6 ASE REPI.SVIIO.N D,"ATE'D

,O./1.6/+,-

HPIB 3.5.2BASESSURVEILLANCE SR 3.5.2.1REQUIREMENTS Verifying the correct alignment for manual and non-automatic poweroperated valves in the HPI flow paths provides assurance that the properflow paths will exist for HPI operation.

This SR does apply to the HPIsuction header cross-connect valves, the HPI discharge cross-connect valves, the HPI discharge crossover valves, and the LPI-HPI flow pathdischarge valves (LP-15 and LP-16). This SR does not apply to valves thatare locked, sealed, or otherwise secured in position, since these valveswere verified to be in the correct position prior to locking,

sealing, orsecuring.

Similarly, this SR does not apply to automatic valves sinceautomatic 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 arein the correct position.

The Surveillance Frequency is based on operatinc>-

experience, equipment reliability, and plant risk and is controlled under [Ithe Surveillance Frequency Control Program.SR 3.5.2.2With 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 todevelop voids and pockets of entrained gases. Venting the HPI pumpcasings periodically reduces the potential that such voids and pockets ofentrained gases can adversely affect operation of the HPI System. Thiswill also reduce the potential for water hammer, pump cavitation, andpumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into thereactor vessel following an ESPS signal. This Surveillance is modified by aNote 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 underthe Surveillance Frequency Control Program.SR 3.5.2.3Periodic 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 arespecified 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 HPIB 3.5.2BASESREFERENCES

1. 10 CFR 50.46. Replace Ref. 5 description with:"ASME Code for Operation and2. UFSAR, Section 15.14.3.3.6.

Maintenance of Nuclear PowerPlants."111 I I U .; V. ./ L~~4. NRC Memorandum to V. Stello, Jr., from R.L. aer,"Recommended Interim Revisions to LCOs f r ECCSComponents,"

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, UnitsNos. 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 HighPressure Injection System Requirements,"

dated December 16,1998.OCONEE UNITS 1, 2, & 3B 3.5.2-14

[BASEG flEVICION BA+ G6/Wi62 1I LPIB 3.5.3BASESSURVEILLANCE SR 3.5.3.2 (continued)

REQUIREMENTS cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, orhydrogen) into the reactor vessel following an ESPS signal or duringshutdown cooling.

This Surveillance is modified by a Note that indicates itis not applicable to operating LPI pump(s).

The Surveillance Frequency is HTbased on operating experience, equipment reliability, and plant risk and iscontrolled under the Surveillance Frequency Control Program.

L/JSR 3.5.3.3Periodic 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 arespecified in the Inser ice Testing Program.

whi'ch Czza", X'Iof the ASME Code. \.SR 3.5.3.4 and SR 3.5.3.5These SRs demonstrate that each automatic LPI valve actuates to therequired position on an actual or simulated ESPS signal and that each LPIpump starts on receipt of an actual or simulated ESPS signal. This SR isnot required for valves that are locked, sealed, or otherwise secured inposition under administrative controls.

The test will be considered satisfactory if control board indication verifies that all components haveresponded to the ESPS actuation signal properly (all appropriate ESPSactuated pump breakers have opened or closed and all ESPS actuatedvalves have completed their travel).

The Surveillance Frequency is basedon operating experience, equipment reliability, and plant risk and iscontrolled 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 3.5.3-8BAGEG REYI 0-611 &1 LPIB 3.5.3BASESSURVEILLANCE REQUIREMENTS (continued)

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

The Surveillance Frequency is based on operating experience, equipment reliability, andplant risk and is controlled under the Surveillance Frequency ControlProgram.REFERENCES

1. 10 CFR 50.46.Replace Ref. 6 description with:"ASME Code for Operation and2. UFSAR, Section 15.14.3.3.6.

Maintenance of Nu3. 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 ECCSComponents,"

December 1, 1975.clear Power6.A&",l Boilo- --A R46664 Vero 'Aifiel Xo6 otoI, Jnoric7. NRC Safety Evaluation of Babcock & Wilcox Owners Group(B&WOG) Topical Report BAW-2295, Revision 1, "Justification forthe Extension of Allowed Outage Time for Low Pressure Injection and Reactor Building Spray systems,"

(TAC No. MA3807) datedJune 30, 1999.OCONEE UNITS 1, 2, & 3B 3.5.3-9C D;AES R'-v"SION i

Reactor Building Spray and Cooling SystemsB 3.6.5BASESSURVEILLANCE 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 ormotor 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 theSurveillance Frequency Control Program.SR 3.6.5.3Verifying that each required Reactor Building Spray pump's developed head at the flow test point is greater than or equal to the requireddeveloped head ensures that spray pump performance has not degradedduring the cycle. Flow and differential pressure are normal tests ofcentrifugal pump performance required (iby Code(Ref. 4). Since the Reactor Building Spray Sys em pumps cannot betested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and isindicative of overall performance.

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

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

Thistest verifies the heat removal capability of the Low Pressure Injection (LPI)Coolers and Reactor Building Cooling Units. The Surveillance Frequency Iis based on operating experience, equipment reliability, and plant risk andIIis 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 SystemsB 3.6.5BASESREFERENCES 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 andMaintenance of Nuclear Power Plants."e.!occsi Codo_', ipostion Xl.L{ppJOCONEE UNITS 1, 2, & 3B 3.6.5-11C 3 .5AiEi RE9'A'RION DTEDW/14/1.,

II MSRVsB 3.7.1BASES (continued)

SURVEILLANCE SR 3.7.1.1REQUIREMENTS This SR verifies the OPERABILITY of the MSRVs by the verification ofMSRV lift setpoints in accordance with the Inservice Testing Pro ram. Thesafety and relief valve tests are performed in accordance withANSICode (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; ande 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 requiredto be performed in MODES 1 and 2. This note allows entry into andoperation in MODE 3 prior to performing the SR, provided there is noevidence that the equipment is otherwise believed to be incapable ofperforming its function.

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

This allows operation to proceed to aMODE or other specified condition where other necessary postmaintenance tests can be completed.

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

The mode change provisions permit entry intoMode 3 in order to test and adjust the set pressure, as necessary, to satisfySR 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 nottested at hot conditions, the lift setting pressure must be corrected toambient 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 ]

MSRVsB 3.7.1BASES (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 andUFSAR, Section 10.3.3. Maintenance of Nuclear Power Plants."10 CFR 50.36.AWI.AA~ S A .19-- PrzzFurz

8. 1 Gods, roctin 'XI.IOCONEE UNITS 1, 2, & 3B 3.7.1-4I AGEG RlEVISION D)ATED $ 061. ij MFCVs and SFCVsB 3.7.3BASES (continued)

SURVEILLANCE REQUIREMENTS SR 3.7.3.1This SR verifies that the closure time of each MFCV and SFCV is_< 25 seconds on an actual or simulated actuation signal. The 25 secondsincludes 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 unitto operation following a refueling outage. The MFCV and SFCV should notbe tested at power since even a part stroke exercise increases the risk of avalve closure with the unit generating power. This is consistent with theASME Coe (t7, Ref. 2) requirements during operation inMODES 1 and 2.This SR is modified by a Note that allows entry into and operation inMODE 3 prior to performing the SR.The Frequency for this SR is in accordance with the Inservice TestingProgram.REFERENCES

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

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

This SRdoes not apply to valves that are locked, sealed, or otherwise secured inposition, since those valves are verified to be in the correct position prior tolocking,

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 anytesting or valve manipulation; rather, it involves verification that thosevalves 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 theSurveillance Frequency Control Program.SR 3.7.5.2Verifying that each EFW pump's developed head at the flow test point isgreater than or equal to the required developed head ensures that EFWpump performance has not degraded below the acceptance criteria duringthe cycle. Flow and dilerential head are ormal indications of pumpperformance required y &e_%@.Xl,4vthe ASME Code (Ref. 3). Becauseit is undesirable to intr uce cold EFW inmt the steam generators whilethey are operating, this test may be performed on a test flow path.This test confirms OPERABILITY, trends performance, and detectsincipient failures by indicating abnormal nerformance.

Performance ofinservice testing in the ASME ode,-&@eet.§.

.V(Ref. 3), at 3 monthintervals, satisfies this requirement.

SR 3.7.5.3This SR verifies that EFW can be delivered to the appropriate steamgenerator in the event of any accident or transient that generates anEmergency Feedwater System initiation signal by demonstrating that eachautomatic valve in the flow path actuates to its correct position on an actualor simulated actuation signal. This SR is not required for valves that arelocked, 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 SystemB 3.7.5BASES (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 andMaintenance of Nuclear Power Plants."OCONEE UNITS 1, 2, & 3B 3.7.5-89 U3A REVS EV',',O, DPATED GS61.I4 License Amendment Request No. 2013-04ATTACHMENT 3Revised Technical Specification Pages[2 pages following this cover page]NOTE: Attached are clean, retyped TS Pages 5.0-12 and -13.

Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.7 Pre-Stressed Concrete Containment Tendon Surveillance Pro-gramThis program provides controls for monitoring any tendon degradation inpre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity.

The program shallinclude baseline measurements prior to initial operations.

The TendonSurveillance

Program, inspection frequencies, and acceptance criteria shall be inaccordance with Section Xl, Subsection IWL of the ASME Boiler and PressureVessel Code and applicable addenda as required by 10 CFR 50.55a, asamended 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 Programinspection frequencies.

5.5.8 Reactor Coolant Pump Flywheel Inspection ProgramThis program shall provide for inspection of each reactor coolant pump flywheel.

At approximately three-year intervals, the bore and keyway of each reactorcoolant pump flywheel shall be subjected to an inplace, volumetric examination.

Whenever maintenance or repair activities necessitate flywheel

removal, asurface examination of exposed surfaces and a complete volumetric examination shall be performed if the interval measured from the previous such inspection isgreater than 6 2/3 years. The interval may be extended up to one year to permitinspections to coincide with a planned outage.5.5.9 Inservice Testinq ProgramThis 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 andMaintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:OCONEE UNITS 1, 2, & 35.0-12Amendment Nos. _, _, & __ I Programs and Manuals5.55.5 Programs and Manuals5.5.9Inservice Testing Program (continued)

ASME OM Code andapplicable Addendaterminology forinservice testingactivities WeeklyMonthlyQuarterly or every3 monthsSemiannually orevery 6 monthsEvery 9 monthsYearly or annuallyBiennially or every 2 yearsRequired Frequencies for performing inservice testing activities At least once per 7 daysAt least once per 31 daysAt least once per 92 daysAt least once per 184 daysAt least once per 276 daysAt least once per 366 daysAt least once per 731 days5.5.10b. The provisions of SR 3.0.2 are applicable to the above requiredFrequencies 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; andd. Nothing in the ASME OM Code shall be construed to supersede therequirements of any TS.Steam Generator (SG) ProgramA Steam Generator Program shall be established and implemented to ensurethat SG tube integrity is maintained.

In addition, the Steam Generator Programshall include the following provisions:

a. Provisions for condition monitoring assessments.

Condition monitoring assessment means an evaluation of the "as found" condition of the tubingwith respect to the performance criteria for structural integrity and accidentinduced leakage.

The "as found" condition refers to the condition of thetubing 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 whichthe SG tubes are inspected or plugged to confirm that the performance criteria are being met.OCONEE UNITS 1, 2, & 35.0-13Amendment Nos. -, -, & I License Amendment Request No. 2013-04ATTACHMENT 4Revised 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 -4B 3.4.14-6B 3.5.2-12 and -14B 3.5.3-8 and -9B 3.6.5-9 and -11B 3.7.1-3 and -4B 3.7.3-4B 3.7.5-6 and -8 Pressurizer Safety ValvesB 3.4.10BASES (continued)

APPLICABLE SAFETY ANALYSESAll accident analyses in the UFSAR that require safety valveactuation assume operation of both pressurizer safety valves to limitincreasing reactor coolant pressure.

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

The startup accident establishes the minimum safetyvalve capacity.

The startup accident is assumed to occur at < 15% power.Single failure of a safety valve is neither assumed in the accident analysisnor required to be addressed by the ASME Code. Compliance with thisSpecification is required to ensure that the accident analysis and designbasis calculations remain valid.Pressurizer safety valves satisfy Criterion 3 of 10 CFR 50.36 (Ref. 3).LCOThe two pressurizer safety valves are set to open at the RCS designpressure (2500 psig) and within the ASME specified tolerance to avoidexceeding the maximum RCS design pressure SL, to maintain accidentanalysis assumptions and to comply with ASME Code requirements.

Thevalves will be tested per ASME Code requirements and returned to servicewith as-left setpoints of 2500 psig +/- 1%. The upper and lower pressuretolerance limits are based on the requirements of the ASME Boiler andPressure Vessel Code,Section III, Article 9, Summer 1967, which limit therise in pressure within the vessel which they protect, to 10% above thedesign pressure.

Inoperability of one or both valves could result inexceeding the SL if a transient were to occur.The consequences of exceeding the ASME pressure limit could includedamage 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 intemperature, OPERABILITY of two valves is required because thecombined capacity is required to keep reactor coolant pressure below110% of its design value during certain accidents.

Portions of MODE 3 areconservatively

included, although the listed accidents may not require bothsafety valves for protection.

OCONEE UNITS 1, 2, & 3B 3.4.10-2xx/xx/xx I

Pressurizer Safety ValvesB 3.4.10BASES (continued)

SURVEILLANCE REQUIREMENTS SR 3.4.10.1SRs are specified in the Inservice Testing Program.

Pressurizer safetyvalves are to be tested in accordance with the requirements of the ASMECode (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. Thesevalues include instrument uncertainties.

REFERENCES

1. ASME, Boiler and Pressure Vessel Code,Section III.2. ASME Code for Operation and Maintenance of Nuclear PowerPlants.3. 10 CFR 50.36.OCONEE UNITS 1, 2, & 3B 3.4.10-4XX/XX/XX I

RCS PIV LeakageB 3.4.14BASESREFERENCES

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

OCONEE UNITS 1, 2, & 3B 3.4.14-6XX/XX/XX I

HPIB 3.5.2BASESSURVEILLANCE SR 3.5.2.1REQUIREMENTS Verifying the correct alignment for manual and non-automatic poweroperated valves in the HPI flow paths provides assurance that the properflow paths will exist for HPI operation.

This SR does apply to the HPIsuction header cross-connect valves, the HPI discharge cross-connect valves, the HPI discharge crossover valves, and the LPI-HPI flow pathdischarge valves (LP-15 and LP-16). This SR does not apply to valves thatare locked, sealed, or otherwise secured in position, since these valveswere verified to be in the correct position prior to locking,

sealing, orsecuring.

Similarly, this SR does not apply to automatic valves sinceautomatic 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 arein the correct position.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled underthe Surveillance Frequency Control Program.SR 3.5.2.2With 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 todevelop voids and pockets of entrained gases. Venting the HPI pumpcasings periodically reduces the potential that such voids and pockets ofentrained gases can adversely affect operation of the HPI System. Thiswill also reduce the potential for water hammer, pump cavitation, andpumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into thereactor vessel following an ESPS signal. This Surveillance is modified by aNote 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 underthe Surveillance Frequency Control Program.SR 3.5.2.3Periodic 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 theInservice Testing Program of the ASME Code.OCONEE UNITS 1, 2, & 3B 3.5.2-12XX/XX/XX I

HPIB 3.5.2BASESREFERENCES

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 ECCSComponents,"

December 1, 1975.5. ASME Code for Operation and Maintenance of Nuclear PowerPlants.6. Letter from R. W. Reid (NRC) to W. 0. Parker, Jr. (Duke)transmitting Safety Evaluation for Oconee Nuclear Station, UnitsNos. 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 HighPressure Injection System Requirements,"

dated December 16,1998.OCONEE UNITS 1, 2, & 3B 3.5.2-14XX/XX/XX I

LPIB 3.5.3BASESSURVEILLANCE SR 3.5.3.2 (continued)

REQUIREMENTS cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, orhydrogen) into the reactor vessel following an ESPS signal or duringshutdown cooling.

This Surveillance is modified by a Note that indicates itis not applicable to operating LPI pump(s).

The Surveillance Frequency isbased on operating experience, equipment reliability, and plant risk and iscontrolled under the Surveillance Frequency Control Program.SR 3.5.3.3Periodic 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 theInservice Testing Program of the ASME Code.SR 3.5.3.4 and SR 3.5.3.5These SRs demonstrate that each automatic LPI valve actuates to therequired position on an actual or simulated ESPS signal and that each LPIpump starts on receipt of an actual or simulated ESPS signal. This SR isnot required for valves that are locked, sealed, or otherwise secured inposition under administrative controls.

The test will be considered satisfactory if control board indication verifies that all components haveresponded to the ESPS actuation signal properly (all appropriate ESPSactuated pump breakers have opened or closed and all ESPS actuatedvalves have completed their travel).

The Surveillance Frequency is basedon operating experience, equipment reliability, and plant risk and iscontrolled 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, & 3B 3.5.3-8XX/XX/XX I

LPIB 3.5.3BASESSURVEILLANCE SR 3.5.3.6REQUIREMENTS (continued)

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

The Surveillance Frequency is based on operating experience, equipment reliability, andplant risk and is controlled under the Surveillance Frequency ControlProgram.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 ECCSComponents,"

December 1, 1975.6. ASME Code for Operation and Maintenance of Nuclear PowerPlants.7. NRC Safety Evaluation of Babcock & Wilcox Owners Group(B&WOG) Topical Report BAW-2295, Revision 1, "Justification forthe Extension of Allowed Outage Time for Low Pressure Injection and Reactor Building Spray systems,"

(TAC No. MA3807) datedJune 30, 1999.OCONEE UNITS 1, 2, & 3B 3.5.3-9XX/XX/XX I

Reactor Building Spray and Cooling SystemsB 3.6.5BASESSURVEILLANCE 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 ormotor 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 theSurveillance Frequency Control Program.SR 3.6.5.3Verifying that each required Reactor Building Spray pump's developed head at the flow test point is greater than or equal to the requireddeveloped head ensures that spray pump performance has not degradedduring the cycle. Flow and differential pressure are normal tests ofcentrifugal pump performance required by the ASME Code (Ref. 4). Sincethe Reactor Building Spray System pumps cannot be tested with flowthrough the spray headers, they are tested on recirculation flow. This testconfirms one point on the pump design curve and is indicative of overallperformance.

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

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

Thistest 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 andis controlled under the Surveillance Frequency Control Program.OCONEE UNITS 1, 2, & 3B 3.6.5-9XX/XX/XX I

Reactor Building Spray and Cooling SystemsB 3.6.5BASESREFERENCES

1. UFSAR, Section 3.1.2. UFSAR, Section 6.2.3. 10 CFR 50.36.4. ASME Code for Operation and Maintenance of Nuclear PowerPlants.OCONEE UNITS 1, 2, & 3B 3.6.5-11XX/XX/XX I

MSRVsB 3.7.1BASES (continued)

SURVEILLANCE SR 3.7.1.1REQUIREMENTS This SR verifies the OPERABILITY of the MSRVs by the verification ofMSRV lift setpoints in accordance with the Inservice Testing Program.

Thesafety 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; ande. Verification of the balancing device integrity on balanced valves.The ASME Code requires the testing of all valves every 5 years, with aminimum of 20% of the valves tested every 24 months.This SR is modified by a Note that states the surveillance is only requiredto be performed in MODES I and 2. This note allows entry into andoperation in MODE 3 prior to performing the SR, provided there is noevidence that the equipment is otherwise believed to be incapable ofperforming its function.

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

This allows operation to proceed to aMODE or other specified condition where other necessary postmaintenance tests can be completed.

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

The mode change provisions permit entry intoMode 3 in order to test and adjust the set pressure, as necessary, to satisfySR 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 nottested at hot conditions, the lift setting pressure must be corrected toambient conditions of the valve at operating temperature and pressure.

OCONEE UNITS 1, 2, & 3B 3.7.1-3XX/XX/XX I

MSRVsB 3.7.1BASES (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 PowerPlants.OCONEE UNITS 1, 2, & 3B 3.7.1-4XX/XX/XX I

MFCVs and SFCVsB 3.7.3BASES (continued)

SURVEILLANCE REQUIREMENTS SR 3.7.3.1This SR verifies that the closure time of each MFCV and SFCV is_< 25 seconds on an actual or simulated actuation signal. The 25 secondsincludes 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 unitto operation following a refueling outage. The MFCV and SFCV should notbe tested at power since even a part stroke exercise increases the risk of avalve closure with the unit generating power. This is consistent with theASME Code (Ref. 2) requirements during operation in MODES 1 and 2.This SR is modified by a Note that allows entry into and operation inMODE 3 prior to performing the SR.The Frequency for this SR is in accordance with the Inservice TestingProgram.REFERENCES

1. 10 CFR 50.36.2. ASME Code for Operation and Maintenance of Nuclear PowerPlants.OCONEE UNITS 1, 2, & 3B 3.7.3-4XX/XX/XX I

EFW SystemB 3.7.5BASES (continued)

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

This SRdoes not apply to valves that are locked, sealed, or otherwise secured inposition, since those valves are verified to be in the correct position prior tolocking,

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 anytesting or valve manipulation; rather, it involves verification that thosevalves 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 theSurveillance Frequency Control Program.SR 3.7.5.2Verifying that each EFW pump's developed head at the flow test point isgreater than or equal to the required developed head ensures that EFWpump performance has not degraded below the acceptance criteria duringthe cycle. Flow and differential head are normal indications of pumpperformance required by the ASME Code (Ref. 3). Because it isundesirable to introduce cold EFW into the steam generators while they areoperating, this test may be performed on a test flow path.This test confirms OPERABILITY, trends performance, and detectsincipient failures by indicating abnormal performance.

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

SR 3.7.5.3This SR verifies that EFW can be delivered to the appropriate steamgenerator in the event of any accident or transient that generates anEmergency Feedwater System initiation signal by demonstrating that eachautomatic valve in the flow path actuates to its correct position on an actualor simulated actuation signal. This SR is not required for valves that arelocked, sealed, or otherwise secured in position under administrative OCONEE UNITS 1, 2, & 3 B 3.7.5-6 XX/XXIXX IlOCONEE UNITS 1, 2, & 3B 3.7.5-6XX/XX/XX I

EFW SystemB 3.7.5BASES (continued)

REFERENCES

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