Text

Duane Arnold - License Amendment Request (TSCR-1 30) for Amendment to Technical Specifications Section 5.5.6 for the Inservice Testing Program
	ML15289A233
Person / Time
Site:	Duane Arnold
Issue date:	10/14/2015
From:	Vehec T A; NextEra Energy Duane Arnold
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NG-15-0284
	Download: ML15289A233 (24)
	v • d • e

NExTeraTEN ERGYNG-1 5-028410 CFR 50.90October 14, 2015U.S. Nuclear Regulatory Commission Attn: Document Control DeskWashington, D.C. 20555-0001 Duane Arnold Energy CenterDocket 50-331License No. DPR-49License Amendment Request (TSCR-1 30) for Amendment to Technical Specifications Section 5.5.6 for the Inservice Testing ProgramReferences

1. TSTF-479-A, Revision 0, "Changes to Reflect Revision of 10 CFR50.55a,"

dated December 19, 20052. TSTF-497, Revision 0, "Limit Inservice Testing Program SR 3.0.2Application to Frequencies of 2 Years or Less," dated July 12, 20063. Letter from T. H. Boyce (USNRC) to members of the Technical Specification Task Force, dated December 6, 20054. Letter from T. J. Kobetz (USNRC) to members of the Technical Specification Task Force, dated October 4, 2006In accordance with the provisions of Section 50.90 of Title 10 of the Code of FederalRegulations (10 CFR), NextEra Energy Duane Arnold, LLC (hereafter, NextEra EnergyDuane Arnold) hereby submits Technical Specification Change Request TSCR-130 torevise Duane Arnold Energy Center (DAEC) Technical Specifications (TS). Specifically, the proposed changes will replace references to Section Xl of the American Society ofMechanical Engineers (ASME) Boiler and Pressure Vessel Code with references to theASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) in TSSection 5.5.6 for the Inservice Testing Program.These proposed changes are based on Technical Specification Task Force (TSTF) 479-A, Revision 0, "Changes to Reflect Revision of 10 CFR 50.55a,"

(Reference

1) asmodified by TSTF-497, Revision 0, "Limit Inservice Testing Program SR 3.0.2Application to Frequencies of 2 Years or Less," (Reference

2) and approved by theNRC in References 3 and 4. These proposed changes will correct or revise TS Section5.5.6 to align with the requirements of 10 CFR 50.55a, "Codes and standards,"

p..q-m:ph (f), "lnservice testing requirements."

In addition; t: thc r *~cmn.e-deletion of the references, NextEra Energy Duane Arnold is also adding a provision inTS Section 5.5.6 to only apply the extension allowance of Surveillance Requirement (SR) 3.0.2 to the frequency table listed in the TS as part of the Inservice TestingProgram and to normal and accelerated inservice testing frequencies of two years orless, as applicable. NextEra Energy Duane Arnold, LLC, 3277 DAEC Road, Palo, IA 52324 Document Control DeskNG-1 5-0284Page 2Attachment 1 provides an evaluation of the proposed changes.

Attachment 2 providesmarked-up pages of existing TS to show the proposed changes.

Attachment 3 providesrevised (clean) TS pages. Attachment 4 provides the marked-up TS Bases pages forinformation only. There are no new Regulatory Commitments or revisions to existingRegulatory Commitments.

Although this request is neither outage related nor required by any specific date,NextEra Energy Duane Arnold requests review and approval of the proposed licenseamendment within one year, with the amendment being implemented within 60 days ofits receipt.In accordance with 10 CFR 50.91l(a)(1),

"Notice for Public Comment,"

the analysisabout the issue of no significant hazards consideration using the standards in 10 CFR50.92 is being provided to the Commission.

In accordance with 10 CFR 50.91(b)(1),

"Notice for Public Comment; StateConsultation,"

a copy of this application and its reasoned analysis about no significant hazards considerations is being provided to .the designated State of Iowa official.

The DAEC Onsite Review Group has reviewed the proposed license amendment request.If you have any questions or require additional information, please contact J. MichaelDavis at 319-851-7032.

I declare under penalty of perjury that the foregoing is true and correct.Exectgdon October 14, 2015.T. A. VehecVice President, Duane Arnold Energy CenterNextEra Energy Duane Arnold, LLCAttachments:

1. Evaluation of Proposed Changes2. Proposed Technical Specification Changes (Mark-up Copy)3. Revised Technical Specification Changes (Clean, Typed)4. Proposed Technical Specification Bases Changes (FYI)cc: Regional Administrator, USNRC, Region Ill,Project Manager, USNRC, Duane Arnold Energy CenterResident Inspector, USNRC, Duane Arnold Energy CenterA. Leek (State of Iowa)

ATTACHMENT 1 TO NG-15-0284 NEXTERA ENERGY DUANE ARNOLD, LLCDUANE ARNOLD ENERGY CENTERLICENSE AMENDMENT REQUEST (TSCR-1 30)FOR AMENDMENT TO TECHNICAL SPECIFICATIONS SECTION 5.5.6 FOR THEINSERVICE TESTING PROGRAMEVALUATION OF PROPOSED CHANGES1.0 DESCRIPTION 2.0 PROPOSED CHANGES3.0 BACKGROUND 4.0 TECHNICAL ANALYSIS5.0 REGULATORY SAFETY ANALYSIS5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements and Criteria6.0 ENVIRONMENTAL CONSIDERATION

7.0 PRECEDENT

8.0 REFERENCES

Page 1 of 8

1.0 DESCRIPTION

The proposed changes will replace references to Section Xl of the American Society ofMechanical Engineers (ASME) Boiler and Pressure Vessel Code with references to theASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) inTechnical Specifications (TS) Section 5.5.6 for the Inservice Testing Program.These proposed changes are based on Technical Specification Task Force (TSTF) 479-A, Revision 0, "Changes to Reflect Revision of 10 CFR 50.55a,"

(Reference

1) asmodified by TSTF-497, Revision 0, "Limit Inservice Testing Program SR 3.0.2Application to Frequencies of 2 Years or Less," (Reference

2) and approved by theNRC in References 3 and 4. These proposed changes will correct or revise TS Section5.5.6 to align with the requirements of 10 CFR 50.55a, "Codes and standards,"

paragraph (f), "lnservice testing requirements."

In addition to the replacement ordeletion of the references, NextEra Energy Duane Arnold is also adding a provision inTS Section 5.5.6 to only apply the extension allowance of Surveillance Requirement (SR) 3.0.2 to the frequency table listed in the TS as part of the Inservice TestingProgram and to normal and accelerated inservice testing frequencies of two years orless, as applicable.

2.0 PROPOSED CHANGESNextEra Energy Duane Arnold proposes to revise the existing wording of TS Section5.5.6 to replace references to the ASME Boiler and Pressure Vessel Code, Section Xlwith references to the ASME OM Code. A marked-up copy of the proposed changes tothe TS is provided in Attachment

2. Attachment 3 provides revised (clean) TS pages.TS Bases Sections will also be revised to replace or delete the references to Section XIof the ASME Boiler and Pressure Vessel Code, as applicable.

Proposed revisions tothe TS Bases are also included for information only in Attachment

4. The changes tothe affected TS Bases pages will be incorporated in accordance with the TS BasesControl Program upon receipt of the NRC approved License Amendment.

3.0 BACKGROUND

In 1990, the ASME published the initial edition of the ASME OM Code that providedrules for inservice testing of pumps and valves. The ASME OM Code replaced SectionXl of the Boiler and Pressure Vessel Code for inservice testing of pumps and valves.The 1995 edition with the 1996 Addenda of the ASME OM Code (Reference

5) wasincorporated by reference into 10 CFR 50.55a paragraph (b) on September 22, 1999(Reference 6). 10 CFR 50.55a paragraph (f), "Inservice testing requirements,"

section(4.)(Ji) requires that inservice testing during successive 120-month intervals comply with.the requirements of the latest edition and addenda of the Code incorporated byreference into 10 CFR 50 .55a(b),

12 months before the start of the 120-month interval.

The ASME OM Code is the Code of record for the current 10-Year inservice testing(IST) Interval for the Duane Arnold Energy Center (DAEC). DAEC currently is in thePage 2 of 8 Fourth 1ST Ten-Year Interval that began on February 1, 2006 and ends on January 31,2016. The ASME OM Code will also be the Code of record for the Fifth IST Ten-YearInterval that begins on February 1, 2016 and ends on January 31, 2026.On February 23, 2006 at a meeting between the TSTF and the NRC, the NRC statedthat they did not agree with the portion of TSTF-479 referring to the application of a 25%IST interval extension for SR 3.0.2 to test frequencies and would not approve plant-specific amendments incorporating that portion of TSTF-479.

Specifically, the NRCexpressed a concern that frequency extensions may be applied to frequencies greaterthan two years and requested that the TSTF be revised to apply the provisions of SR3.0.2 to the table listed in the TS as part of the Inservice Testing Program and to normaland accelerated inservice testing frequencies of two years or less. The NRC stated thatthey would accept applying SR 3.0.2 to IST Frequencies not listed in the Inservice Testing Program table provided that those Frequencies are specified in the Inservice Testing Program as 2 years or less.On July 12, 2006, TSTF-497, Revision 0, (Reference

2) was submitted to reflect therevised NRC position

.These proposed changes to TS Section 5.5.6 are based onTSTF 479-A, Revision 0, as modified by TSTF-497, Revision 0, which was approved bythe NRC on October 4, 2006 (Reference 4).4.0 TECHNICAL ANALYSISOn September 22, 1999, the NRC amended 10 CFR 50.55a, "Codes and Standards,"

by Final Rule (64 FR 51370) (Reference

6) to incorporate by reference more recenteditions and addenda of the ASME Boiler and Pressure Vessel Code and the ASMEOM Code for construction, inservice inspection, and inservice testing of thosecomponents.

The 2001 edition and the 2002 and 2003 Addenda of the ASME OM Code wasapproved for use by the NRC and was incorporated by reference into 10 CFR 50.55aparagraph (b) on October 1, 2004 (Reference 7).The ASME OM Code is the Code of record for the current 10-Year IST Interval forDAEC. TS Section 5.5.6 currently references the ASME Boiler and Pressure VesselCode, Section Xl, as the standard for testing frequencies and inservice testing of ASMECode Class 1, 2, and 3 pumps and valves. The proposed changes to TS Section 5.5.6will replace references to Section Xl of the ASME Boiler and Pressure Vessel Code withreferences to the ASME OM Code as applicable to meet the requirements of 10 CFR50.55a(f)(4),

as amended in Reference 7.5.0 REGULATORY SAFETY ANALYSIS

-~5.1 No Significant Hazards Consideration NextEra Energy Duane Arnold has evaluated the proposed changes to theTechnical Specifications (TS) using the criteria in 10 CFR 50.92 and hasPage 3 of 8 determined that the proposed changes do not involve a significant hazardsconsideration.

Description of Amendment Request:

The requested amendment would modifythe TS by replacing references to Section Xl of the American Society ofMechanical Engineers (ASME) Boiler and Pressure Vessel Code with references to the ASME Code for Operation and Maintenance of Nuclear Power Plants (OMCode) in TS Section 5.5.6 for the Inservice Testing Program.These proposed changes are based on Technical Specification Task Force(TSTF) 479-A, Revision 0, "Changes to Reflect Revision of 10 CFR 50.55a,"(Reference

1) as modified by TSTF-497, Revision 0, "Limit Inservice TestingProgram SR 3.0.2 Application to Frequencies of 2 Years or Less," (Reference 2)and approved by the NRC in References 3 and 4. These proposed changes willcorrect or revise TS Section 5.5.6 to align with the requirements of 10 CFR50.55a, "Codes and standards,"

paragraph (f), "lnservice testing requirements."

In addition to the replacement or deletion of the references, NextEra EnergyDuane Arnold is also adding a provision in TS Section 5.5.6 to only apply theextension allowance of Surveillance Requirement (SR) 3.0.2 to the frequency table listed in the TS as part of the Inservice Testing Program and to normal andaccelerated inservice testing frequencies of two years or less, as applicable.

Basis for proposed no significant hazards determination:

As required by 10 CFR50.91(a),

the NextEra Energy Duane Arnold analysis of the issue of no significant hazards consideration is presented below:1. Does the proposed change involve a significant increase in the probability orconsequences of an accident previously evaluated?

Response:

NoThe proposed changes revise TS Section 5.5.6 to conform to the requirements of10 CFR 50.55a, "Codes and standards,"

paragraph (f) regarding the inservice testing of pumps and valves. TS Section 5.5.6 currently references the ASMEBoiler and Pressure Vessel Code, Section Xl, requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The proposedchanges would reference the ASME GM Code as applicable, which is consistent with 10 CFR 50.55a, paragraph (f), "lnservice testing requirements."

In addition, the proposed changes clarify that the extension allowance of SR 3.0.2 onlyapplies to the frequency table listed in the TS, if applicable, as part of theInservice Testing Program and to normal and accelerated inservice testingfrequencies of two years or less. The definitions of the frequencies are notchanged by the requested amendment.

The proposed changes are administrative in nature, do not affect any accidentinitiators, do not affect the ability to successfully respond to previously evaluated accidents and do not affect radiological assumptions used in the evaluations.

Page 4 of 8 Thus, the probability or radiological consequences of any accident previously evaluated are not increased.

Therefore, the proposed changes do not involve a significant increase in theprobability or consequences of an accident previously evaluated.

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

Response:

NoThe proposed changes revise TS Section 5.5.6 to conform to the requirements of10 CFR 50.55a(f) regarding the inservice testing .of pumps and valves. TSSection 5.5.6 Currently references the ASME Boiler and Pressure Vessel Code,Section Xl, requirements for the inservice testing of ASME Code Class 1, 2, and3 pumps and valves. The proposed changes would reference the ASME OMCode as applicable, which is consistent with 10 CFR 50.55a(f).

In addition, theproposed changes clarify that the extension allowance of SIR 3.0.2 only applies tothe frequency table listed in the TS, if applicable, as part of the Inservice TestingProgram and to normal and accelerated inservice testing frequencies of twoyears or less. The definitions of the frequencies are not changed by therequested amendment.

The proposed changes to TS Section 5.5.6 do not affect the performance of anystructure, system, or component credited with mitigating any accident previously evaluated and do not introduce any new modes of system operation or failuremechanisms.

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

3. Does the proposed change involve a significant reduction in a margin ofsafety?Response:

NoMargin of safety is related to confidence in the ability of the fission productbarriers (fuel cladding, reactor coolant system, and primary containment) toperform their design functions during and following postulated accidents.

Theproposed changes do not affect the function of the reactor coolant pressureboundary or its response during plant transients.

The proposed changes reviseTS Section 5.5.6 to conform to the requirements of 10 CFR 50.55a(f) regarding the inservice testing of pumps and valves.TS Section 5.5.6 currently references the ASME Boiler and Pressure VesselCode, Section Xl, requirements for the inservice testing of ASME Code Class 1,2, and 3 pumps and valves. The proposed changes would reference the ASMEPage 5 of 8 GM Code as applicable, which is consistent with 10 CFR 50.55a(f).

In addition, the proposed changes clarify that the extension allowance of Surveillance Requirement (SR) 3.0.2 only applies to the frequency table listed in the TS, ifapplicable, as part of the Inservice Testing Program and to normal andaccelerated inservice testing frequencies of two years or less. The definitions ofthe frequencies are not changed by the requested amendment.

The proposed changes do not alter the manner in which safety limits, limitingsafety system settings or limiting conditions for operation are determined.

Thesafety analysis acceptance criteria are not affected by this change. Theproposed change will not result in plant operation in a configuration outside thedesign basis. The proposed change does not adversely affect systems thatrespond to safely shutdown the plant and to maintain the plant in a safeshutdown condition.

Therefore, the proposed changes do not involve a significant reduction in amargin of safety.Based on the above, NextEra Energy Duane Arnold concludes that the proposedchanges present no significant hazards consideration under the standards setforth in 10 CFR 50.92(c),

and, accordingly, a finding of"no significant hazardsconsideration" is justified.

5.2 Applicable Regulatory Requirements and Criteria10 CFR 50.55a defines the requirements for applying industry Codes to alicensed boiling or pressurized water-cooled nuclear power facility.

10 CFR50.55a(f)(4) requires that throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves which are classified asASME Code Class 1, Class 2, and Class 3 must meet the inservice testrequirements that are incorporated by reference in 10 CFR 50.55a(b) to theextent practical within the limitations of design, geometry and materials ofconstruction of the components.

10 CFR 50.55a(f)(4)(ii) further states that inservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during successive 120-month intervals must comply with the latest edition andaddenda of the Code, incorporated by reference in 10 CFR 50.55a(b),

12 monthsbefore the start of the 120-month interval.

10 CFR 50.55a(f)(5)(ii) states that if a revised inservice test program for a facilityconflicts with the TS for the facility, the licensee shall apply to the NRC for -amendment of the TS to conform the TS to the revised program.

This application shall be submitted at least six months before the start of the period during whichthe provisions become applicable.

Page 6 of 8 NextEra Energy Duane Arnold has identified that implementation of the DAECFourth IST Ten-Year Interval Program does not reflect the requirements specified in TS Section 5.5.6. Therefore, in accordance with the requirements of 10 CFR50.55a(f)(5)(ii),

NextEra Energy Duane Arnold is submitting this LicenseAmendment Request to correct this administrative oversight.

6.0 ENVIRONMENTAL CONSIDERATION 10 CFR 51 .22(c)(9) provides criteria for and identification of licensing andregulatory actions eligible for categorical exclusion from performing anenvironmental assessment.

A proposed amendment of an operating license fora facility requires no environmental assessment, if the operation of the facility inaccordance with the proposed amendment does not: (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released

offsite, or (3) resultin a significant increase in individual or cumulative occupational radiation exposure.

NextEra has reviewed this license amendment request anddetermined that the proposed amendment meets the eligibility criteria forcategorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR51.22(b),

no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the amendment.

Thebasis for this determination is as follows.BasisThis change meets the eligibility criteria for categorical exclusion set forth in 10CFR 51 .22(c)(9) for the following reasons:..........

1.-AS-demonsra-ted in- the-10 CFR 50.92 evaluation, the proposedamendment does not involve a significant hazards consideration.

2. The proposed amendment does not result in a significant change in thetypes or significant increase in the amounts of any effluents that may bereleased offsite.

The proposed amendment does not change or modifythe design or operation of any plant systems, structures, or components.

The proposed amendment does not affect the amount or types ofgaseous, liquid, or solid waste generated onsite. The proposedamendment does not directly or indirectly affect effluent discharges.

3. The proposed amendment does not result in a significant increase inindividual or cumulative occupational radiation exposure.

The proposedamendment does not change or modify the design or operation of anyplant systems, structures, or components.

The proposed amendment does not directly or indirectly affect the radiological source terms.Page 7 of 8

7.0 PRECEDENT

This License Amendment Request is similar to a License Amendment Requestapproved by letter dated August 28, 2008 (Reference 8).

8.0 REFERENCES

1. TSTF-479-A, Revision 0, "Changes to Reflect Revision of 10 CFR50.55a,"

dated December 19, 20052. TSTF-497, Revision 0, "Limit lnservice Testing Program SR 3.0.2Application to Frequencies of 2 Years or Less," dated July 12, 20063. Letter from T. H. Boyce (USNRC) to members of the Technical Specification Task Force, dated December 6, 20054. Letter from T. J. Kobetz (USNRC) to members of the Technical Specification Task Force, dated October 4, 20065. American Society of Mechanical Engineers (ASME), "Operation andMaintenance of Nuclear Power Plants (OM Code)," 1995 Edition throughthe 1996 Addenda6. Federal Register, Volume 64, Number 183, "10 CFR Part 50 -IndustryCodes and Standards; Amended Requirements,"

dated September 22,1999.....7_--Feder~l-Re~liS-ter, Volume 69, Number 190, "10 CFR Part 50 -IndustryCodes and Standards; Amended Requirements,"

dated October 1, 20048. Letter from C Gratton (USNRC) to C. G. Pardee (Exelon),

"Braidwood

Station, Units 1 and 2; Byron Station, Unit Nos. 1 and 2; Dresden NuclearPower Station, Units 2 and 3; Limerick Generating

Station, Units 1 and 2;Oyster Creek Nuclear Generating Station; Peach Bottom Atomic PowerStation, 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 TravelerTSTF-479 and TSTF-497 (TAC NOS. MD6530 THRU MD6543),"

datedAugust 28, 2008 (ML080600330)

Page 8 of 8 ATTACHMENT 2 TO NG-15-0284 NEXTERA ENERGY DUANE ARNOLD, LLCDUANE ARNOLD ENERGY CENTERLICENSE AMENDMENT REQUEST (TSCR-1 30)FOR AMENDMENT TO TECHNICAL SPECIFICATIONS SECTION 5.5.6 FOR THEINSERVICE TESTING PROGRAMPROPOSED TECHNICAL SPECIFICATIONS CHANGES(MARKUP COPY)1 page follows Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.6Inservice Testinq Programapplicable to theASME Code forOperations andMaintenance ofNuclear Power Plants(ASME OM Code)This program provides controls for inservice testing of ASME Code Class1, 2, and 3 components.

The program shall include the following:

a.Tesin cp1cfidin Section X ofA thI ..I oiefollows:aria appilicale MaUUelaa are asASME Boiler and Proc'urc-Vessee-Code andapplicable Addendaterminology for ReqLinservice testing for pquired Frequencies erforming inservice nn WeeklyMonthlyBiquarterly Quarterly or every3 monthsSemiannually orevery 6 monthsEvery 9 monthsYearly or annuallyBiennially or everyAt least once perAt least once perAt least once per7 days31 days46 daysAt least once per 92 daysand to other normaland accelerated Frequencies specified as 2 years or less inthe Inservice TestingProgramAt least once perAt least once perAt least once per184 days276 days366 days2Years At least once per 731 daysb. Tep sos of SR 3.0.2 are applicable to the above requiredFeunifor performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testingactivities; andd. Nothing in the ASME Bceicr end,-P""....+ -Vee .e, Code shall beconstrued to supersede the requirements of any TS.(continued)

DAEC5.0-11DAE 5.-11Amendment No. 2.,:4 ATTACHMENT 3 TO NG-15-0284 NEXTERA ENERGY DUANE ARNOLD, LLCDUANE ARNOLD ENERGY CENTERLICENSE AMENDMENT REQUEST (TSCR-1 30)FOR AMENDMENT TO TECHNICAL SPECIFICATIONS SECTION 5.5.6 FOR THEINSERVICE TESTING PROGRAMREVISED TECHNICAL SPECIFICATIONS PAGES1 page follows Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.6Inservice Testinqi ProgqramThis program provides controls for inservice testing of ASME Code Class1, 2, and 3 components.

The program shall include the following:

a. Testing Frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) andapplicable Addenda are as follows:ASME GM Code andapplicable Addendaterminology forinservice testingactivities Required Frequencies for performing inservice testing activities WeeklyMonthlyBiquarterly Quarterly or every3 monthsSemiannually orevery 6 monthsEvery 9 monthsYearly or annuallyBiennially or every2 yearsAt least once per 7 daysAt least once per 31 daysAt least once per 46 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 and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program forperforming inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testingactivities; andd. Nothing in the ASME GM Code shall be construed to supersede the requirements of any TS.(continued)

DAEC5.0-11DAEC .0-11Amendment No.

ATTACHMENT 4 TO NG-15-0284 NEXTERA ENERGY DUANE ARNOLD, LLCDUANE ARNOLD ENERGY CENTERLICENSE AMENDMENT REQUEST (TSCR-1 30)FOR AMENDMENT TO TECHNICAL SPECIFICATIONS SECTION 5.5.6 FOR THEINSERVICE TESTING PROGRAMPROPOSED TECHNICAL SPECIFICATION BASES CHANGES(FOR INFORMATION ONLY)9 pages follow SRVs and SVsB 3.4.3B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.3 Safety Relief Valves (SRVs) and Safety Valves (SVs)BASESBACKGROUND The ASME Boi!or an~d Proc'-re Code requires the reactorpressure vessel be protected from overpressure during upsetconditions by self-actuated safety valves. As part of the nuclearpressure relief system, the size and number of SRVs and SVs areselected such that peak pressure in the nuclear system will notexceed the ASME Code limits for the Reactor Coolant PressureBoundary (RCPB).The SRVs and SVs are located on the main steam lines betweenthe reactor vessel and the first isolation valve within the drywell.The SRVs can actuate by either of two modes: the safety mode orthe relief mode. However, for the purpose of this LCO, only thesafety mode is required.

The SVs actuate only in the safety mode.In the safety mode (or spring mode of operation),

the spring loadedpilot valve opens when steam pressure at the valve inletovercomes the spring force holding the pilot valve closed. Openingthe pilot valve allows a pressure differential to develop across themain valve piston and opens the main valve. The safety modefunction of both SRVs and SVs satisfies the Code requirement.

Apower generation design basis function of the SRVs is also toprevent opening of the SVs during normal plant isolations and loadrejections.

Each SRV discharges steam through a discharge line to a pointbelow the minimum water level in the suppression pool while theSVs discharge directly to the drywell airspace.

The SRVs thatprovide the relief mode are the Low-Low Set (LLS) valves and theAutomatic Depressurization System (ADS) valves. The LLSrequirements are specified in LCO 3.6.1.5, "Low-Low Set (LLS)Valves,"

and the ADS requirements are specified in LCO 3.5.1,"ECCS -- Operating."

APPLICABLE The overpressure protection system must accommodate the mostSAFETY severe pressurization transient.

Evaluations have determined thatANALYSES the most severe transient is the closure of all Main SteamIsolation Valves (MSIVs),

followed by reactor scram on highneutron flux (i.e., failure of the direct scram associated with MSIVposition)

(Ref. 1). For the purpose of the analyses, 6 valves (anycombination of SRVs and SVs) are assumed to operate in the(continued)

DAECDACB 3.4-15 ITC-10 h,Amcndmcn~t 223 SRVs and SVsB 3.4.3BASES (continued)

SURVEILLANCE SR 3.4.3.1REQUIREMENTS This Surveillance requires that the SRVs and SVs will open at thepressures assumed in the safety analysis of Reference

1. Thedemonstration of the SRV and SV lift settings must be performed during shutdown, since this is a bench test, to be done inaccordance with the Inservice Testing Program.

The lift settingpressure shall correspond to ambient conditions of the valves atnominal operating temperatures and pressures.

The SRV and SVsetpoints are +/- 3% for OPERABILITY; however the valves arereset to +/- 1% during the Surveillance to allow for drift.The Surveillance Frequency is in accordance with the Inservice Testing Program requirements contained in the ASME Code7This Surveillance must be performed during shutdownconditions.

SR 3.4.3.2The actuator of each dual function safety/relief valves (S/RVs) isstroked to verify that the pilot valve strokes when manuallyactuated.

The actuator test is performed by energizing a solenoidthat pneumatically actuates a plunger.

The plunger is connected to the second stage disc located within the main valve body.When steam pressure actuates the plunger during plant operation, this allows pressure to be vented from the top of the main valvepiston, allowing reactor pressure to lift the main valve piston,which opens the main valve disc. The test will verify movement ofthe plunger in accordance with vendor recommendations.

However, since this test is performed prior to establishing thereactor pressure needed to overcome main valve closure forces,the main valve disc will not stroke during the test.This SR, together with the valve testing performed as required bythe ASME Code for pressure relieving devices (ASME OM Code -2001 through 2003 Addenda),

verify the capability of each reliefvalve to perform its function.

Valve testing will be performed at a steam test facility, where thevalve (i.e., main valve and pilot valve) and an actuatorrepresentative of the actuator used at the plant will be installed ona steam header in the same orientation as the plant installation.

The test conditions in the test facility will be similar to those in theplant installation, including ambient temperature, valve insulation, and steam conditions.

The valve will then be leak tested,functionally tested to ensure the valve is capable of opening and(continued)

DAEC B 3.4-19 TSC R-12--j SRVs and SVsB 3.4.3BASESSURVEILLANCE SR 3.4.3.2 (continued)

REQUIREMENTS closing (including stroke time), and leak tested a final time. Valveseat tightness will be verified by a cold bar test, and if not free offog, leakage will be measured and verified to be below designlimits. In addition, for the safety mode of S/RVs, an as-foundsetpoint verification and as-found leak check are performed, followed by verification of set pressure, and delay time. The valvewill then be shipped to the plant without any disassembly oralteration of the main valve or pilot valve components.

The combination of the valve testing and the valve actuator testingprovide a complete check of the capability of the valves to openand close, such that full functionality is demonstrated throughoverlapping tests, without cycling the valves.If a valve fails to actuate due only to the failure of the solenoid butis capable of opening on overpressure, the safety function of theSRV is not considered inoperable.

This SR is not applicable to the SVs, due to their design whichdoes not include the manual relief capability, nor do they have adischarge line that can become blocked.The Frequency of this SR is in accordance with the Inservice Testing Program.REFERENCES

1. UFSAR, Section 5.2.2.2.1.IASME Code for Operation andjMaintenance of Nuclear Power PlantsI2. UFSAR, Section 15.1.2.4. NUREG 1482, Guidelines for Inservice Testing at NuclearPower Plants.ETDAECB 3.4-20TSCR-4a8 EGGS -Operating B 3.5.1BASESSURVEILLANCE SR 3.5.1.3REQU IREMENTS(continued)

Verification that a 100 day supply of nitrogen exists for each ADS Jaccumulator ensures adequate nitrogen pressure for reliable ADSoperation.

The accumulator on each ADS valve providespneumatic pressure for valve actuation.

The design pneumatic supply pressure requirements for the accumulator are such thatfollowing a failure of the pneumatic supply to the accumulator, each ADS valve can be actuated at least 5 times up to 100 daysfollowing a LOGA (Reference 4). This SR can be met by either: 1)verifying that the drywell nitrogen header supply pressure is > 90psig, or 2) when drywell nitrogen header supply pressure is < 90psig, using the actual accumulator check valve leakage ratesobtained from the most-recent tests to determine, analytically, thata 100 day supply of nitrogen exists for each accumulator.

Theresults of this analysis can also be used to determine when the100 day supply of nitrogen will no longer exist for individual ADSaccumulators, and when each ADS valve would subsequently berequired to be declared inoperable, assuming the drywell nitrogensupply pressure is not restored to > 90 psig. The Surveillance Frequency is controlled under the Surveillance Frequency ControlProgram.

The Frequency takes into consideration administrative controls over operation of the nitrogen system and alarms for lownitrogen pressure.

SR 3.5.1.4.

SR 3.5.1.5, and SR 3.5.1.6The performance requirements of the low pressure EGGS pumpsare determined through application of the 10 GFR 50, Appendix Kcriteria (Ref. 8). This periodic Surveillance is performed (inaccordance with the ASME Code~v-.,,,et"i"

..,,v' requirements for theEGGS pumps) to verify that the EGGS pumps will develop the flowrates required by the respective analyses.

The low pressureEGGS pump flow rates ensure that adequate core cooling isprovided to satisfy the acceptance criteria of Reference

10. Thepump flow rates are verified against a system head equivalent tothe RPV pressure expected during a LOGA. The total systempump outlet pressure is adequate to overcome the elevation headpressure between the pump suction and the vessel discharge, thepiping friction losses, and RPV pressure present during a LOCA.These values may be established during preoperational testing orby analysis.

(continued)

DAEC B 3.5-15 TSC R--12-0 LLS ValvesB 3.6.1.5BASESSURVEILLANCE SR 3.6.1.5.1 (continued)

REQUIREMENTS limits. In addition, for the safety mode of S/RVs, an as-found setpointverification and as-found leak check are performed, followed byverification of set pressure, and delay time. The valve will then beshipped to the plant without any disassembly or alteration of the mainvalve or pilot valve components.

The combination of the valve testing and the valve actuator testingprovide a complete check of the capability of the valves to open andclose, such that full functionality is demonstrated through overlapping tests, without cycling the valves.The Frequency of this SR is in accordance with the Inservice TestingProgram.SR 3.6.1.5.2 The LLS designated SRVs are required to actuate automatically uponreceipt of specific initiation signals.

A system functional test isperformed to verify that the mechanical portions (i.e., solenoids) of theLLS function operate as designed when initiated either by an actual orsimulated automatic initiation signal. The LOGIC SYSTEMFUNCTIONAL TEST in LCO 3.3.6.3, "Low-Low Set (LLS)Instrumentation,"

overlaps this SR to provide complete testing of thesafety function.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The Frequency is based on the need toperform this Surveillance under the conditions that apply during a plantoutage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience hasshown these components usually pass the Surveillance when performed at this Frequency.

Therefore, the Frequency was concluded to beacceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation.

Thisprevents a reactor pressure vessel pressure blowdown.

REFERENCES

1. UFSAR, Section 5.4.13 Maintenance of Nuclear Power Plants2. ASME, Boilor an"d Proccu'ro Vocscl Codo, Soctionq XI.3. NEDE-30021-P, Low-Low Set Relief Logic System and LowerMSlV Water Level Trip for DAEC, January 1983.7JDAECB 3.6-36TSCR-42-8 RHR Suppression Pool CoolingB 3.6.2.3BASES (continued)

SURVEILLANCE SR 3.6.2.3.1 REQUIREMENTS Verifying by administrative means the correct alignment formanual, power operated and automatic valves in the RHRsuppression pool cooling mode flow path provides assurance thatthe proper flow path exists for system operation.

This SR doesnot apply to valves that are locked, sealed, or otherwise securedin position since these valves were verified to be in the correctposition prior to locking, sealing or securing.

A valve is alsoallowed to be in the nonaccident position provided it can bealigned to the accident position within the time assumed in theaccident analysis.

This is acceptable since the RHR suppression pool cooling mode is manually initiated.

This SR does not requireany testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correctposition.

This SR does not apply to manual valves or to valvesthat cannot be inadvertently misaligned, such as check valves.The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The Frequency is justified becausethe valves are operated under procedural

control, improper valveposition would affect only a single subsystem, the probability of anevent requiring initiation of the system is low, and the subsystem is a manually initiated system. This Frequency has been shownto be acceptable based on operating experience.

SR 3.6.2.3.2 Verifying that each RHR pump develops a flow rate > 4800 gpmwhile operating in the suppression pool cooling mode with flowthrough the associated heat exchanger ensures that the primarycontainment peak pressure and temperature and the localsuppression pool temperature can be maintained below designlimits. This test also verifies that pump performance has notdegraded during the surveillance interval.

Flow is a normal test ofcentrifugal pump performance required by ASME CodeT8eetien-4 (Ref. 2). This test confirms one point on the pumpdesign curve, and the results are indicative of overallperformance.

Such inservice testing confirms component OPERABILITY, trends performance, and detects incipient failuresby indicating abnormal performance.

The Frequency of this SR isin accordance with the Inservice Testing Program.SR 3.6.2.3.3 RHR Suppression Pool Cooling System piping and components (continued)_

DAEC B 3.6-63 TSCR-44.6 RHR Suppression Pool CoolingB 3.6.2.3BASESSURVEILLANCE REQUIREMENTS SR 3.6.2.3.3 (continued) accumulated gas void volume has been evaluated anddetermined to not challenge system OPERABILITY.

The accuracyof the method used for monitoring the susceptible locations andtrending of the results should be sufficient to assure systemOPERABILITY during the Surveillance interval.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The Surveillance Frequency mayvary by location susceptible to gas accumulation.

REFERENCES lASME Code for Operation and1. UFSAR, Section 15.2.1.1.

Maintenance of Nuclear Power Plantsf-.DAECB 3.6-64DAECB 3.-64TSCR-1446 AC Sources -Operating B 3.8.1BASESSURVEILLANCE REQUIREMENTS (continued)

SR 3.8.1.6This Surveillance demonstrates that each required fuel oil transferpump operates and transfers fuel oil from its associated storagetank to its associated day tank. It is required to supportcontinuous operation of standby power sources.

ThisSurveillance provides assurance that the fuel oil transfer pump isOPERABLE, the fuel oil piping system is intact, the fuel deliverypiping is not obstructed, and the controls and control systems formanual fuel transfer systems are OPERABLE.

Additional assurance of fuel oil transfer pump OPERABILITY is provided bymeeting the testing requirements for pumps that are contained inthe ASME Boilcr and Precssure Vessel Code, ectie,",X4 (Ref. 13).The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.I SR 3.8.1.7See SR 3.8.1.2.SR 3.8.1.8The slow transfer of each 4.16 kV essential bus power supplyfrom the preferred offsite circuit (i.e. -the startup transformer) tothe alternate preferred offsite circuit (i.e. the standby transformer) demonstrates the OPERABILITY of the alternate preferred circuitdistribution network to power the shutdown loads. TheSurveillance Frequency is controlled under the Surveillance Frequency Control Program.

The Frequency of the Surveillance is based on engineering judgment taking into consideration theplant conditions required to perform the Surveillance, and isintended to be consistent with expected fuel cycle lengths.Operating experience has shown that these components usuallypass the SR when performed on this Frequency.

Therefore, theFrequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note. The reason for the Note is that,during operation with the reactor critical, performance of this SRcould cause perturbations to the Electrical Distribution Systemsthat could challenge continued steady state operation and, as aresult, plant safety systems.

Credit may be taken for unplanned events that satisfy this SR.(continued)_

DAECB 3.8-19TSCR--1-20 AC Sources -Operating B 3.8.1BASESREFERENCES (continued) 6.7.8.9.10.11.12.13.14.15.16.17.Regulatory Guide 1.93.Generic Letter 84-15.UFSAR, Section 3.1.2.2.9 Regulatory Guide 1.108.Regulatory Guide 1.137.IASME Code for Operation and[Deleted]

IMaintenance of Nuclear Power PlantsUFSAR, Section 15.2.1 Boi,,,r ..nd PrDe......

Vccc .. o '"", Soction XI.IEEE Standard 308.[Deleted]

UFSAR, Table 8.3-1.Regulatory Guide 1.9.TSCR-082.

DAECDAEC B 3.8-25 NExTeraTEN ERGYNG-1 5-028410 CFR 50.90October 14, 2015U.S. Nuclear Regulatory Commission Attn: Document Control DeskWashington, D.C. 20555-0001 Duane Arnold Energy CenterDocket 50-331License No. DPR-49License Amendment Request (TSCR-1 30) for Amendment to Technical Specifications Section 5.5.6 for the Inservice Testing ProgramReferences