Text

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; NextEra Energy Duane Arnold
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NG-15-0284
	Download: ML15289A233 (24)
	v • d • e

NExTera EN ERGY T October 14, 2015 NG-1 5-0284 10 CFR 50.90 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Duane Arnold Energy Center Docket 50-331 License No. DPR-49 License Amendment Request (TSCR-1 30) for Amendment to Technical Specifications Section 5.5.6 for the Inservice Testing Program References : 1. TSTF-479-A, Revision 0, "Changes to Reflect Revision of 10 CFR 50.55a," dated December 19, 2005

2. TSTF-497, Revision 0, "Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less," dated July 12, 2006

3. Letter from T. H. Boyce (USNRC) to members of the Technical Specification Task Force, dated December 6, 2005

4. Letter from T. J. Kobetz (USNRC) to members of the Technical Specification Task Force, dated October 4, 2006 In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), NextEra Energy Duane Arnold, LLC (hereafter, NextEra Energy Duane Arnold) hereby submits Technical Specification Change Request TSCR-130 to revise Duane Arnold Energy Center (DAEC) Technical Specifications (TS). Specifically, the proposed changes will replace references to Section Xl of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code with references to the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 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 Testing Program 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 will correct or revise TS Section 5.5.6 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 in TS 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 Testing Program and to normal and accelerated inservice testing frequencies of two years or less, as applicable. ****

NextEra Energy Duane Arnold, LLC, 3277 DAEC Road, Palo, IA 52324

Document Control Desk NG-1 5-0284 Page 2 provides an evaluation of the proposed changes. Attachment 2 provides marked-up pages of existing TS to show the proposed changes. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides the marked-up TS Bases pages for information only. There are no new Regulatory Commitments or revisions to existing Regulatory 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 license amendment within one year, with the amendment being implemented within 60 days of its receipt.

In accordance with 10 CFR 50.91l(a)(1), "Notice for Public Comment," the analysis about the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is being provided to the Commission.

In accordance with 10 CFR 50.91(b)(1), "Notice for Public Comment; State Consultation," 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. Michael Davis at 319-851-7032.

I declare under penalty of perjury that the foregoing is true and correct.

Exectgdon October 14, 2015.

T. A. Vehec Vice President, Duane Arnold Energy Center NextEra Energy Duane Arnold, LLC Attachments: 1. Evaluation of Proposed Changes

2. 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 Center Resident Inspector, USNRC, Duane Arnold Energy Center A. Leek (State of Iowa)

ATTACHMENT 1 TO NG-15-0284 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER LICENSE AMENDMENT REQUEST (TSCR-1 30)

FOR AMENDMENT TO TECHNICAL SPECIFICATIONS SECTION 5.5.6 FOR THE INSERVICE TESTING PROGRAM EVALUATION OF PROPOSED CHANGES

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

S

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements and Criteria

6.0 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 of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code with references to the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) in Technical 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) as modified by TSTF-497, Revision 0, "Limit Inservice Testing Program 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 will correct or revise TS Section 5.5.6 to align with the requirements of 10 CFR 50.55a, "Codes and standards,"

paragraph (f), "lnservice testing requirements." In addition to the replacement or deletion of the references, NextEra Energy Duane Arnold is also adding a provision in TS 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 Testing Program and to normal and accelerated inservice testing frequencies of two years or less, as applicable.

2.0 PROPOSED CHANGE

S NextEra Energy Duane Arnold proposes to revise the existing wording of TS Section 5.5.6 to replace references to the ASME Boiler and Pressure Vessel Code, Section Xl with references to the ASME OM Code. A marked-up copy of the proposed changes to the 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 XI of the ASME Boiler and Pressure Vessel Code, as applicable. Proposed revisions to the TS Bases are also included for information only in Attachment 4. The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control 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 provided rules for inservice testing of pumps and valves. The ASME OM Code replaced Section Xl 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) was incorporated 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 by reference 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 the Page 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-Year Interval 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 stated that 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 NRC expressed a concern that frequency extensions may be applied to frequencies greater than two years and requested that the TSTF be revised to apply the provisions of SR 3.0.2 to the table listed in the TS as part of the Inservice Testing Program and to normal and accelerated inservice testing frequencies of two years or less. The NRC stated that they 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 the revised NRC position . These proposed changes to TS Section 5.5.6 are based on TSTF 479-A, Revision 0, as modified by TSTF-497, Revision 0, which was approved by the NRC on October 4, 2006 (Reference 4).

4.0 TECHNICAL ANALYSIS

On 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 recent editions and addenda of the ASME Boiler and Pressure Vessel Code and the ASME OM Code for construction, inservice inspection, and inservice testing of those components.

The 2001 edition and the 2002 and 2003 Addenda of the ASME OM Code was approved for use by the NRC and was incorporated by reference into 10 CFR 50.55a paragraph (b) on October 1, 2004 (Reference 7).

The ASME OM Code is the Code of record for the current 10-Year IST Interval for DAEC. TS Section 5.5.6 currently references the ASME Boiler and Pressure Vessel Code, Section Xl, as the standard for testing frequencies and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes to TS Section 5.5.6 will replace references to Section Xl of the ASME Boiler and Pressure Vessel Code with references to the ASME OM Code as applicable to meet the requirements of 10 CFR 50.55a(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 the Technical Specifications (TS) using the criteria in 10 CFR 50.92 and has Page 3 of 8

determined that the proposed changes do not involve a significant hazards consideration.

Description of Amendment Request: The requested amendment would modify the TS by replacing references to Section Xl of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code with references to the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 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 Testing Program 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 will correct or revise TS Section 5.5.6 to align with the requirements of 10 CFR 50.55a, "Codes and standards," paragraph (f), "lnservice testing requirements."

In addition to the replacement or deletion of the references, NextEra Energy Duane Arnold is also adding a provision in TS 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 Testing Program and to normal and accelerated inservice testing frequencies of two years or less, as applicable.

Basis for proposed no significant hazards determination: As required by 10 CFR 50.91(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 or consequences of an accident previously evaluated?

Response: No The proposed changes revise TS Section 5.5.6 to conform to the requirements of 10 CFR 50.55a, "Codes and standards," paragraph (f) regarding the inservice testing of pumps and valves. TS Section 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, and 3 pumps and valves. The proposed changes 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 only applies to the frequency table listed in the TS, if applicable, as part of the Inservice Testing Program and to normal and accelerated inservice testing frequencies of two years or less. The definitions of the frequencies are not changed by the requested amendment.

The proposed changes are administrative in nature, do not affect any accident initiators, 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 the probability or consequences of an accident previously evaluated.

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

Response: No The proposed changes revise TS 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 Vessel Code, Section Xl, requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes would reference the ASME OM Code as applicable, which is consistent with 10 CFR 50.55a(f). In addition, the proposed changes clarify that the extension allowance of SIR 3.0.2 only applies to the frequency table listed in the TS, if applicable, as part of the Inservice Testing Program and to normal and accelerated inservice testing frequencies of two years or less. The definitions of the frequencies are not changed by the requested amendment.

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

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

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

Response: No Margin of safety is related to confidence in the ability of the fission product barriers (fuel cladding, reactor coolant system, and primary containment) to perform their design functions during and following postulated accidents. The proposed changes do not affect the function of the reactor coolant pressure boundary or its response during plant transients. The proposed changes revise TS 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 Vessel Code, Section Xl, requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes would reference the ASME Page 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, if applicable, as part of the Inservice Testing Program and to normal and accelerated inservice testing frequencies of two years or less. The definitions of the frequencies are not changed by the requested amendment.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed change will not result in plant operation in a configuration outside the design basis. The proposed change does not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition.

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

Based on the above, NextEra Energy Duane Arnold concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of"no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements and Criteria 10 CFR 50.55a defines the requirements for applying industry Codes to a licensed boiling or pressurized water-cooled nuclear power facility. 10 CFR 50.55a(f)(4) requires that throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves which are classified as ASME Code Class 1, Class 2, and Class 3 must meet the inservice test requirements that are incorporated by reference in 10 CFR 50.55a(b) to the extent practical within the limitations of design, geometry and materials of construction 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 and addenda of the Code, incorporated by reference in 10 CFR 50.55a(b), 12 months before the start of the 120-month interval.

10 CFR 50.55a(f)(5)(ii) states that if a revised inservice test program for a facility conflicts 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 which the provisions become applicable.

Page 6 of 8

NextEra Energy Duane Arnold has identified that implementation of the DAEC Fourth 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 CFR 50.55a(f)(5)(ii), NextEra Energy Duane Arnold is submitting this License Amendment Request to correct this administrative oversight.

6.0 ENVIRONMENTAL CONSIDERATION

10 CFR 51 .22(c)(9) provides criteria for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment of an operating license for a facility requires no environmental assessment, if the operation of the facility in accordance 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) result in a significant increase in individual or cumulative occupational radiation exposure. NextEra has reviewed this license amendment request and determined that the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the amendment. The basis for this determination is as follows.

Basis This change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51 .22(c)(9) for the following reasons:

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

2. The proposed amendment does not result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite. The proposed amendment does not change or modify the design or operation of any plant systems, structures, or components.

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

3. The proposed amendment does not result in a significant increase in individual or cumulative occupational radiation exposure. The proposed amendment does not change or modify the design or operation of any plant 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 Request approved by letter dated August 28, 2008 (Reference 8).

8.0 REFERENCES

1. TSTF-479-A, Revision 0, "Changes to Reflect Revision of 10 CFR 50.55a," dated December 19, 2005

2. TSTF-497, Revision 0, "Limit lnservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less," dated July 12, 2006

3. Letter from T. H. Boyce (USNRC) to members of the Technical Specification Task Force, dated December 6, 2005

4. Letter from T. J. Kobetz (USNRC) to members of the Technical Specification Task Force, dated October 4, 2006

5. American Society of Mechanical Engineers (ASME), "Operation and Maintenance of Nuclear Power Plants (OM Code)," 1995 Edition through the 1996 Addenda

6. FederalRegister, Volume 64, Number 183, "10 CFR Part 50 - Industry Codes and Standards; Amended Requirements," dated September 22, 1999

..... 7_--Feder~l-Re~liS-ter, Volume 69, Number 190, "10 CFR Part 50 - Industry Codes and Standards; Amended Requirements," dated October 1, 2004

8. Letter from C Gratton (USNRC) to C. G. Pardee (Exelon), "Braidwood Station, Units 1 and 2; Byron Station, Unit Nos. 1 and 2; Dresden Nuclear Power Station, Units 2 and 3; Limerick Generating Station, Units 1 and 2; Oyster Creek Nuclear Generating Station; Peach Bottom Atomic Power Station, Units 2 and 3; Quad Cities Nuclear Power Station, Units 1 and 2; and Three Mile Island Nuclear Station, Unit 1 - Issuance of Amendments that Adopt Technical Specification Task Force (TSTF) Change Traveler TSTF-479 and TSTF-497 (TAC NOS. MD6530 THRU MD6543)," dated August 28, 2008 (ML080600330)

Page 8 of 8

ATTACHMENT 2 TO NG-15-0284 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER LICENSE AMENDMENT REQUEST (TSCR-1 30)

FOR AMENDMENT TO TECHNICAL SPECIFICATIONS SECTION 5.5.6 FOR THE INSERVICE TESTING PROGRAM PROPOSED TECHNICAL SPECIFICATIONS CHANGES (MARKUP COPY) 1 page follows

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.6 Inservice Testinq Program This program provides controls for inservice testing of ASME Code Class applicable to the 1, 2, and 3 components. The program shall include the following:

ASME Code for Operations and a. Tesin Frequenc...Ies*

cp1cfidin Section X ofA thI I .. oie Maintenance of aria appilicale MaUUelaa are as follows:

Nuclear Power Plants (ASME OM Code) ASME Boiler and Proc'urc

-Vessee-Code and applicable Addenda terminology for ReqL uired Frequencies inservice testing for pqerforming inservice nn nir~tiviti*~

Weekly At least once per 7 days Monthly At least once per 31 days Biquarterly At least once per 46 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days and to other normal Every 9 months At least once per 276 days and accelerated Yearly or annually At least once per 366 days Frequencies specified Biennially or every as 2 years or less in the Inservice Testing Program b. Tep 2Years sos of SR 3.0.2 are applicable At least onceabove to the per 731 days required Feunifor performing inservice testing activities;

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

d. Nothing in the ASME Bceicr end,-P""....+ -

Vee .e, Code shall be construed to supersede the requirements of any TS.

(continued)

DAEC 5.0-11 DAE 5.-11Amendment No. 2.,:4

ATTACHMENT 3 TO NG-15-0284 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER LICENSE AMENDMENT REQUEST (TSCR-1 30)

FOR AMENDMENT TO TECHNICAL SPECIFICATIONS SECTION 5.5.6 FOR THE INSERVICE TESTING PROGRAM REVISED TECHNICAL SPECIFICATIONS PAGES 1 page follows

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.6 Inservice Testinqi Progqram This program provides controls for inservice testing of ASME Code Class 1, 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) and applicable Addenda are as follows:

ASME GM Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Biquarterly At least once per 46 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

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

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

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

(continued)

DAEC 5.0-11 DAEC .0-11Amendment No.

ATTACHMENT 4 TO NG-15-0284 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER LICENSE AMENDMENT REQUEST (TSCR-1 30)

FOR AMENDMENT TO TECHNICAL SPECIFICATIONS SECTION 5.5.6 FOR THE INSERVICE TESTING PROGRAM PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (FOR INFORMATION ONLY) 9 pages follow

SRVs and SVs B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 Safety Relief Valves (SRVs) and Safety Valves (SVs)

BASES BACKGROUND The ASME Boi!or an~d Proc'-re V*ccol* Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of SRVs and SVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the Reactor Coolant Pressure Boundary (RCPB).

The SRVs and SVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell.

The SRVs can actuate by either of two modes: the safety mode or the relief mode. However, for the purpose of this LCO, only the safety mode is required. The SVs actuate only in the safety mode.

In the safety mode (or spring mode of operation), the spring loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve. The safety mode function of both SRVs and SVs satisfies the Code requirement. A power generation design basis function of the SRVs is also to prevent opening of the SVs during normal plant isolations and load rejections.

Each SRV discharges steam through a discharge line to a point below the minimum water level in the suppression pool while the SVs discharge directly to the drywell airspace. The SRVs that provide the relief mode are the Low-Low Set (LLS) valves and the Automatic Depressurization System (ADS) valves. The LLS requirements 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 most SAFETY severe pressurization transient. Evaluations have determined that ANALYSES the most severe transient is the closure of all Main Steam Isolation Valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 1). For the purpose of the analyses, 6 valves (any combination of SRVs and SVs) are assumed to operate in the (continued)

DAEC DACB3.4-15 ITC-10 h,Amcndmcn~t 223

SRVs and SVs B 3.4.3 BASES (continued)

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillance requires that the SRVs and SVs will open at the pressures assumed in the safety analysis of Reference 1. The demonstration of the SRV and SV lift settings must be performed during shutdown, since this is a bench test, to be done in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The SRV and SV setpoints are +/- 3% for OPERABILITY; however the valves are reset 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 Code7

e~eRX. This Surveillance must be performed during shutdown conditions.

SR 3.4.3.2 The actuator of each dual function safety/relief valves (S/RVs) is stroked to verify that the pilot valve strokes when manually actuated. The actuator test is performed by energizing a solenoid that 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 valve piston, allowing reactor pressure to lift the main valve piston, which opens the main valve disc. The test will verify movement of the plunger in accordance with vendor recommendations.

However, since this test is performed prior to establishing the reactor 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 by the ASME Code for pressure relieving devices (ASME OM Code -

2001 through 2003 Addenda), verify the capability of each relief valve to perform its function.

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

The test conditions in the test facility will be similar to those in the plant 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 -j

SRVs and SVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.2 (continued)

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

The combination of the valve testing and the valve actuator testing provide a complete check of the capability of the valves to open and close, such that full functionality is demonstrated through overlapping tests, without cycling the valves.

If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the SRV is not considered inoperable.

This SR is not applicable to the SVs, due to their design which does not include the manual relief capability, nor do they have a discharge 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 and jMaintenance of Nuclear Power PlantsI

2. UFSAR, Section 15.1.2.

4. NUREG 1482, Guidelines for Inservice Testing at Nuclear Power Plants.

ET DAEC B 3.4-20 TSCR-4a8

EGGS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.3 REQU IREMENTS (continued)

Verification that a 100 day supply of nitrogen exists for each ADS J accumulator ensures adequate nitrogen pressure for reliable ADS operation. The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The design pneumatic supply pressure requirements for the accumulator are such that following a failure of the pneumatic supply to the accumulator, each ADS valve can be actuated at least 5 times up to 100 days following a LOGA (Reference 4). This SR can be met by either: 1) verifying that the drywell nitrogen header supply pressure is > 90 psig, or 2) when drywell nitrogen header supply pressure is < 90 psig, using the actual accumulator check valve leakage rates obtained from the most-recent tests to determine, analytically, that a 100 day supply of nitrogen exists for each accumulator. The results of this analysis can also be used to determine when the 100 day supply of nitrogen will no longer exist for individual ADS accumulators, and when each ADS valve would subsequently be required to be declared inoperable, assuming the drywell nitrogen supply pressure is not restored to > 90 psig. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency takes into consideration administrative controls over operation of the nitrogen system and alarms for low nitrogen pressure.

SR 3.5.1.4. SR 3.5.1.5, and SR 3.5.1.6 The performance requirements of the low pressure EGGS pumps are determined through application of the 10 GFR 50, Appendix K criteria (Ref. 8). This periodic Surveillance is performed (in accordance with the ASME Code~v-.,,,et"i" .. ,,v'requirements for the EGGS pumps) to verify that the EGGS pumps will develop the flow rates required by the respective analyses. The low pressure EGGS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of Reference 10. The pump flow rates are verified against a system head equivalent to the RPV pressure expected during a LOGA. The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during a LOCA.

These values may be established during preoperational testing or by analysis.

(continued)

DAEC B 3.5-15 TSC R--12-0

LLS Valves B 3.6.1.5 BASES SURVEILLANCE SR 3.6.1.5.1 (continued)

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

The combination of the valve testing and the valve actuator testing provide a complete check of the capability of the valves to open and close, such that full functionality is demonstrated through overlapping tests, without cycling the valves.

The Frequency of this SR is in accordance with the Inservice Testing Program.

SR 3.6.1.5.2 The LLS designated SRVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to verify that the mechanical portions (i.e., solenoids) of the LLS function operate as designed when initiated either by an actual or simulated automatic initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.3, "Low-Low Set (LLS)

Instrumentation," overlaps this SR to provide complete testing of the safety function.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient ifthe Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at this Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation. This prevents a reactor pressure vessel pressure blowdown.

REFERENCES 1. UFSAR, Section 5.4.13 Maintenance of Nuclear Power Plants

2. ASME, Boilor an"d Proccu'ro Vocscl Codo, Soctionq XI.

3. NEDE-30021-P, Low-Low Set Relief Logic System and Lower MSlV Water Level Trip for DAEC, January 1983. 7 J DAEC B 3.6-36 TSCR-42-8

RHR Suppression Pool Cooling B 3.6.2.3 BASES (continued)

SURVEILLANCE SR 3.6.2.3.1 REQUIREMENTS Verifying by administrative means the correct alignment for manual, power operated and automatic valves in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to manual valves or to valves that cannot be inadvertently misaligned, such as check valves.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. This Frequency has been shown to be acceptable based on operating experience.

SR 3.6.2.3.2 Verifying that each RHR pump develops a flow rate > 4800 gpm while operating in the suppression pool cooling mode with flow through the associated heat exchanger ensures that the primary containment peak pressure and temperature and the local suppression pool temperature can be maintained below design limits. This test also verifies that pump performance has not degraded during the surveillance interval. Flow is a normal test of centrifugal pump performance required by ASME CodeT 8eetien-4 (Ref. 2). This test confirms one point on the pump design curve, and the results are indicative of overall performance. Such inservice testing confirms component OPERABILITY, trends performance, and detects incipient failures by indicating abnormal performance. The Frequency of this SR is in 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 Cooling B 3.6.2.3 BASES SURVEILLANCE SR 3.6.2.3.3 (continued)

REQUIREMENTS accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

REFERENCES 1. UFSAR, Section 15.2.1.1. lASME Maintenance of Nuclear Code for Power Operation and Plants f-.

DAEC B 3.6-64 DAECB 3.-64TSCR-1446

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.6 REQUIREMENTS (continued) This Surveillance demonstrates that each required fuel oil transfer pump operates and transfers fuel oil from its associated storage tank to its associated day tank. It is required to support continuous operation of standby power sources. This Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for manual fuel transfer systems are OPERABLE. Additional assurance of fuel oil transfer pump OPERABILITY is provided by meeting the testing requirements for pumps that are contained in the ASME Boilcr and Precssure Vessel Code, ectie,",X4 (Ref. 13).

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.I SR 3.8.1.7 See SR 3.8.1.2.

SR 3.8.1.8 The slow transfer of each 4.16 kV essential bus power supply from the preferred offsite circuit (i.e. - the startup transformer) to the alternate preferred offsite circuit (i.e. the standby transformer) demonstrates the OPERABILITY of the alternate preferred circuit distribution network to power the shutdown loads. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency of the Surveillance is based on engineering judgment taking into consideration the plant conditions required to perform the Surveillance, and is intended to be consistent with expected fuel cycle lengths.

Operating experience has shown that these components usually pass the SR when performed on this Frequency. Therefore, the Frequency 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 SR could cause perturbations to the Electrical Distribution Systems that could challenge continued steady state operation and, as a result, plant safety systems. Credit may be taken for unplanned events that satisfy this SR.

(continued)_

DAEC B 3.8-19 TSCR--1-20

AC Sources - Operating B 3.8.1 BASES REFERENCES 6. Regulatory Guide 1.93.

(continued)

7. Generic Letter 84-15.

8. UFSAR, Section 3.1.2.2.9

9. Regulatory Guide 1.108.

10. Regulatory Guide 1.137.

IASME Code for Operation and

11. [Deleted] IMaintenance of Nuclear Power Plants

12. UFSAR, Section 15.2.1 *

13. ^SE*A Boi,,,r

.. nd PrDe...... Vccc

.. '"",o Soction XI.

14. IEEE Standard 308.

15. [Deleted]

16. UFSAR, Table 8.3-1.

17. Regulatory Guide 1.9.

DAEC DAEC B 3.8-25 TSCR-082.