Text

License Amendment Request and Exemption Request to Support the Implementation of Framatome High Thermal Performance Fuel
	ML18187A417
Person / Time
Site:	Palo Verde
Issue date:	07/06/2018
From:	Lacal M; Arizona Public Service Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML18190A206	List: ML18187A417;
References
	102-07727-MLL/SMM
	Download: ML18187A417 (364)
	v • d • e

Enclosure Attachments 10 and 11 contain PROPRIETARY information To be withheld under 10 CFR 2.390 10 CFR 50.90 10 CFR 50.12 MARIA L. LACAL Senior Vice President Nuclear Regulatory and Oversight Palo Verde Nuclear Generating Station P.O. Box 52034 Phoenix, AZ 85072 Mail Station 7605 102-07727-MLL/SMM Tel 623.393.6491 July 6, 2018 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Dear Sirs:

Subject:

Palo Verde Nuclear Generating Station Units 1, 2, and 3 Docket Nos. STN 50-528,59-529, and 50-530 License Amendment Request and Exemption Request to Support the Implementation of Framatome High Thermal Performance Fuel In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Arizona Public Service (APS) is submitting a request for an amendment of the Palo Verde Nuclear Generating Station (PVNGS) Operating License Nos. NPF-41, NPF-51, and NPF-74 by revising the Technical Specifications (TS) to support the implementation of Framatome (formerly AREVA, Inc.) Advanced Combustion Engineering 16x16 (CE-16)

High Thermal Performance (HTP') fuel design with M5 as a fuel rod cladding material and gadolinia as a burnable absorber. In addition to the license amendment, pursuant to 10 CFR 50.12, Specific Exemptions, APS is requesting an exemption from the requirements of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, and 10 CFR 50, Appendix K, ECCS Evaluation Models, to allow the use of the Framatome M5 alloy as a fuel rod cladding material.

The enclosure to this letter provides a description and assessment of the proposed changes including a technical evaluation, a regulatory evaluation, a no significant hazards consideration, and an environmental evaluation. The A member of the STARS Alliance, LLC Callaway

Diablo Canyon

Palo Verde

Wolf Creek Attachments 10 and 11 transmitted herewith contain PROPRIETARY information.

When separated from Attachments 10 and 11, this transmittal is decontrolled.

102-07727-MLL/SMM ATTN: Document Control Desk U.S. Nuclear Regulatory Commission License Amendment Request for Implementation of Framatome Fuel Page 2 enclosure also includes eleven attachments. Attachment 1 provides a new regulatory commitment (as defined by NEI 99-04, Guidelines for Managing NRC Commitment Changes, Revision 0) to be implemented. Attachment 2 provides marked-up existing TS pages. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides marked-up TS Bases pages to show the conforming changes for information only. Attachment 5 provides an assessment of limitations and conditions contained in the safety evaluations for NRC-approved topical reports related to this license amendment request. contains two affidavits signed by Framatome that set forth the basis on which the proprietary information in Attachment 11 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4). Correspondence with respect to the proprietary aspects of Attachment 11 or the supporting Framatome affidavits should be addressed to Philip A. Opsal of Framatome. is an affidavit signed by APS that sets forth the basis on which the proprietary information in Attachment 10 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4). Correspondence with respect to the proprietary aspects of Attachment 10 or the supporting APS affidavit should be addressed to Bruce Rash of APS. provides a non-proprietary version of the technical analysis supporting this LAR. Attachment 10 provides a proprietary version of the technical analysis supporting this LAR, which contains information proprietary to APS. provides a non-proprietary version of the Framatome licensing summary reports for large and small break loss of coolant accidents (LOCAs) supporting this LAR. Attachment 11 provides proprietary versions of these reports supporting this LAR, which contains information proprietary to Framatome.

A pre-submittal meeting was held with the NRC on April 5, 2018 (Agency Document Access and Management System [ADAMS] Accession Number ML181028212) to discuss various aspects of this LAR.

In accordance with the PVNGS Quality Assurance Program Description, the Plant Review Board has reviewed and approved the proposed amendment.

By copy of this letter, this license amendment request is being forwarded to

102-07727-MLL/SMM ATTN: Document Control Desk U.S. Nuclear Regulatory Commission License Amendment Request for Implementation of Framatome Fuel Page 3 the Arizona Department of Health Services Bureau of Radiation Control in accordance with 10 CFR 50.91(b)(1).

APS requests approval of the proposed license amendment within 12 months following acceptance of this LAR to support planned implementation during the Spring 2020 Unit 2 refueling outage. APS will implement the approved license amendment within 120 days.

Should you have any questions concerning the content of this letter, please contact Matthew S. Cox, Licensing Section Leader, at (623) 393-5753.

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

Executed on ______July 6, 2018______

(Date)

Sincerely, Digitally signed by Lacal, Lacal, Maria Maria L(Z06149)

DN: cn=Lacal, Maria L(Z06149)

L(Z06149) Date: 2018.07.06 13:45:34

-07'00' MLL/MSC/SMM/maw

Enclosure:

Description and Assessment of Proposed License Amendment cc: K. M. Kennedy NRC Region IV Regional Administrator M. D. Orenak NRC NRR Project Manager for PVNGS M. M. OBanion NRC NRR Project Manager C. A. Peabody NRC Senior Resident Inspector for PVNGS T. Morales Arizona Department of Health Services - Bureau of Radiation Control (ADHS)

ENCLOSURE Description and Assessment of Proposed License Amendment

Enclosure Description and Assessment of Proposed License Amendment TABLE OF CONTENTS

1.

SUMMARY

DESCRIPTION .................................................................................................... 2

2. PROPOSED CHANGES TO PVNGS LICENSING BASIS ............................................................ 2

3. TECHNICAL SPECIFICATIONS CHANGES .............................................................................. 2 3.1. TS 2.1.1.2 - Peak Fuel Centerline Temperature Safety Limit ........................................... 3 3.2. TS 4.2.1 - Design Features (Fuel Assemblies) .................................................................. 3 3.3. TS 5.6.5.b - Core Operating Limits Report (COLR) Analytical Methods........................... 4

4. REGULATORY ANALYSIS FOR TECHNICAL SPECIFICATIONS CHANGES ............................... 5 4.1. Applicable Regulatory Requirements............................................................................... 5 4.2. Precedent ..................................................................................................................... .... 6 4.3. Regulatory Discussion of Topical Reports ........................................................................ 6 4.4. No Significant Hazards Consideration Determination ..................................................... 9

5. ENVIRONMENTAL EVALUATION ....................................................................................... 11

6. PERMANENT EXEMPTION - 10 CFR 50.46 AND 10 CFR PART 50 APPENDIX K FOR M5 .. 11

7. REFERENCES ...................................................................................................................... 14 ATTACHMENTS:

ATTACHMENT 1 - Regulatory Commitment ATTACHMENT 2 - Technical Specifications Page Mark-ups ATTACHMENT 3 - Clean Technical Specifications Pages ATTACHMENT 4 - Technical Specification Bases Page Mark-ups (provided for information only)

ATTACHMENT 5 - Assessment of Topical Report Limitations and Conditions ATTACHMENT 6 - Affidavits from Framatome Submitted in Accordance with 10 CFR 2.390 to Consider Attachment 11 as a Proprietary Document ATTACHMENT 7 - Affidavit from Arizona Public Service Company Submitted in Accordance with 10 CFR 2.390 to Consider Attachment 10 as a Proprietary Document ATTACHMENT 8 - Technical Analysis [NON-PROPRIETARY VERSION]

ATTACHMENT 9 - Framatome Licensing Summary Reports [NON-PROPRIETARY VERSIONS]

ATTACHMENT 10 - Technical Analysis [PROPRIETARY VERSION]

ATTACHMENT 11 - Framatome Licensing Summary Reports [PROPRIETARY VERSIONS]

1

Enclosure Description and Assessment of Proposed License Amendment

1.

SUMMARY

DESCRIPTION This evaluation supports a request to amend Operating Licenses NPF-41, NPF-51, and NPF-74, for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3, respectively.

The license amendment request (LAR) will revise the Technical Specifications (TS) for PVNGS Units 1, 2, and 3 to support implementation of Framatome (formerly AREVA) Advanced Combustion Engineering 16x16 (CE-16) High Thermal Performance (HTP') fuel design with M5 as a fuel rod cladding material and gadolinia as a burnable absorber. This LAR will adapt the approved PVNGS reload analysis methodology to address both Westinghouse and Framatome fuel, including the implementation of Framatome methodologies, parameters and correlations. The ability to use either Westinghouse or Framatome fuel will ensure security of the PVNGS fuel supply by providing for multiple fuel vendors with reliable fuel designs and geographically diverse manufacturing facilities.

In addition to this license amendment request, PVNGS is requesting an exemption from certain requirements of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, and 10 CFR 50, Appendix K, ECCS Evaluation Models, to allow the use of the Framatome M5 alloy as a fuel rod cladding material.

This change is planned to be first implemented with the Spring 2020 Unit 2 refueling outage.

2. PROPOSED CHANGES TO PVNGS LICENSING BASIS Implementation of Framatome fuel at PVNGS Units 1, 2 and 3 requires the following changes to the PVNGS licensing basis:

TS 2.1.1.2 - Clarify that the existing TS is a Westinghouse-supplied fuel peak fuel centerline temperature safety limit and add a Framatome-supplied peak fuel centerline temperature safety limit.

TS 4.2.1 - Simplify the list of fuel rod cladding materials to zirconium-alloy clad, add a paragraph break prior to the text addressing lead test assemblies and delete the last sentence which duplicates the requirements in 10 CFR 50.12 within the TS.

TS 5.6.5 - Add to the listing of analytical methods used to determine the core operating limits as described in documents previously reviewed and approved by the NRC.

A permanent exemption from certain requirements of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, and 10 CFR 50 Appendix K, ECCS Evaluation Models.

3. TECHNICAL SPECIFICATIONS CHANGES The proposed implementation of Framatome fuel requires changes to TS 2.1.1.2, TS 4.2.1, and TS 5.6.5. The following subsections address these changes to the PVNGS licensing basis.

Mark-ups of the affected TS pages are provided in Attachment 2 to this enclosure. Clean TS pages are provided in Attachment 3 to this enclosure.

2

Enclosure Description and Assessment of Proposed License Amendment Mark-ups of the affected Technical Specification Bases pages are provided in Attachment 4 for information in support of the TS changes.

3.1. TS 2.1.1.2 - Peak Fuel Centerline Temperature Safety Limit The restrictions of TS 2.1.1.2 prevent overheating of the fuel and cladding and possible cladding perforation that would result in the release of fission products to the reactor coolant. The current TS 2.1.1.2 text is consistent with the Standard Technical Specifications for Combustion Engineering Plants as specified in NUREG-1432 (References 3 and 4).

Centerline fuel melt analyses for Framatome supplied fuel will be performed with COPERNIC best-estimate predictions and nominal fuel rod design parameters consistent with the best estimate fuel temperature relationship, including uncertainty, defined by Equation 12-3 in approved COPERNIC Topical Report BAW-10231(P)(A) (Reference 10).

The proposed change will revise the PVNGS Units 1, 2, and 3 TS to clarify that the current TS 2.1.1.2 is applicable to Westinghouse supplied fuel. The proposed change will also revise the PVNGS Units 1, 2, and 3 TS to specify a peak fuel centerline temperature safety limit for Framatome supplied fuel that is consistent with Equation 12-3 in approved COPERNIC Topical Report BAW-10231(P)(A). Equation 12-3 is presented in units of degrees Centigrade; for consistency with the existing relationship for Westinghouse supplied fuel, this relationship is incorporated into the TS with units of degrees Fahrenheit.

3.2. TS 4.2.1 - Design Features (Fuel Assemblies)

Technical Specification 4.2.1, Fuel Assemblies, provides a list of approved fuel rod cladding materials, including Zircaloy, ZIRLO and Optimized ZIRLO'. Optimized ZIRLO' was previously added to this list following approval of a permanent exemption request from the requirements of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors and 10 CFR 50 Appendix K, ECCS Evaluation Models.

The proposed change will revise the PVNGS Units 1, 2, and 3 TS to allow the use of M5 fuel rod cladding material. This change will be implemented by simplifying the TS 4.2.1 list of fuel rod cladding materials to the phrase zirconium-alloy clad rather than continuing to individually list each type of cladding. The phrase zirconium-alloy is a generic description that covers the fuel that is licensed at PVNGS.

The proposed change will also revise the PVNGS Units 1, 2, and 3 TS to add a paragraph break prior to the sentence beginning A limited number of lead test assemblies.. This paragraph break is appropriate because the use of lead test assemblies is not related to the TS 4.2.1 text characterizing the fuel assemblies.

The proposed change will also revise the PVNGS Units 1, 2, and 3 TS to delete the TS 4.2.1 sentence requiring an approved exemption of use of other cladding material. This sentence is unnecessary because it does not impose any requirement that is not already imposed by 10 CFR 50.46. In addition, the removal of this sentence is consistent with the NUREG-1432, Standard Technical Specifications, Combustion Engineering Plants (Reference 3).

3

Enclosure Description and Assessment of Proposed License Amendment 3.3. TS 5.6.5.b - Core Operating Limits Report (COLR) Analytical Methods Technical Specification 5.6.5.b, Core Operating Limits Report (COLR), lists the documents which describe the COLR analytical methods used to determine the core operating limits presented in each PVNGS unit-specific COLR.

The proposed change will revise TS 5.6.5.b by updating the references to be consistent with the analytical methods that will be used to determine the core operating limits following Framatome fuel implementation. The proposed change will add the following NRC approved topical reports to the list of referenced analytical methods for consistency with the analytical methods that will be used to determine the core operating limits following Framatome fuel, critical heat flux (CHF) correlation, and gadolinia burnable absorber methodology implementation:

EMF-2103P-A, Realistic Large Break LOCA Methodology for Pressurized Water Reactors

EMF-2328(P)(A), PWR Small Break LOCA Evaluation Model, S-RELAP5 Based

BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code

BAW-10241(P)(A), BHTP DNB Correlation Applied with LYNXT

EPRI-NP-2511-CCM-A, VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores to this enclosure provides an assessment as to how the conditions and limitations contained in the NRC safety evaluations (SEs) for these topical reports are met.

Topical report CENPD-183-A, Loss of Flow, C-E Methods for Loss of Flow Analysis, is currently addressed in TS 5.6.5.b as Item 19. Section 6 of Attachment 5 to this enclosure addresses how topical report CENPD-183-A will be implemented with the introduction of Framatome HTP' fuel.

Clean TS pages are provided in Attachment 3. PVNGS has adopted Technical Specification Task Force (TSTF) 363, Revise Topical Report References in ITS 5.6.5, COLR, (Reference 6) in Amendment No. 137 to the PVNGS Operating Licenses (Reference 18); therefore, the proposed change to TS 5.6.5.b, identifies the documents by number and title.

Each PVNGS unit-specific COLR specifies the complete identification (i.e., TS reference number, title, report number, revision, date, and any supplements) for each of the TS referenced topical reports used to prepare the COLR. The following table describes the complete identification for the additional proposed COLR reference additions to TS 5.6.5.b. Each PVNGS unit-specific COLR will be updated as part of the implementation of the approved license amendment.

4

Enclosure Description and Assessment of Proposed License Amendment Proposed TS Title Report No. Rev Date Suppl Ref No.

Realistic Large Break LOCA Methodology 27 EMF-2103P-A 3 June 2016 N.A.

for Pressurized Water Reactors PWR Small Break LOCA Evaluation Model, 28 EMF-2328(P)(A) 0 March 2001 N.A.

S-RELAP5 Based PWR Small Break LOCA Evaluation Model, December 28 EMF-2328(P)(A) 0 1(P)(A)

S-RELAP5 Based 2016 COPERNIC Fuel Rod Design Computer 29 BAW-10231P-A 1 January 2004 N.A.

Code 30 BHTP DNB Correlation Applied with LYNXT BAW-10241(P)(A) 1 July 2005 N.A.

Volume 1-4 (February VIPRE-01: A Thermal Hydraulic Analysis EPRI-NP-2511-31 Mod 02 2017), N.A.

Code for Reactor Cores CCM-A Volume 5 (March 1988)

4. REGULATORY ANALYSIS FOR TECHNICAL SPECIFICATIONS CHANGES 4.1. Applicable Regulatory Requirements The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.

Title 10 of the Code of Federal Regulations (10 CFR) Paragraph 50.36(c)(2)(ii) requires that TS limiting conditions for operation be established for process variables, design features, or operating restrictions for which a value is assumed as an initial condition of a design basis accident or transient analysis in the licensee's safety analyses. To eliminate the need for an amendment to update the cycle-specific parameter limits for each fuel cycle while complying with 10 CFR 50.36(c)(2)(ii) requirements, the cycle-specific parameter limits are incorporated in the PVNGS unit-specific COLR.

The proposed change adds to TS 5.6.5.b several fuel vendor topical reports which have been reviewed and approved by the NRC for licensing application. The addition of these topical reports to the PVNGS Units 1, 2, and 3 TS is consistent with the current TS 5.6.5.b practice established in Amendment 137 to the PVNGS Operating Licenses (Reference 6, Section 2.5) to identify the documents by number and title. The proposed change to the PVNGS unit-specific COLR specifies the complete identification (i.e., technical specification reference number, title, report number, revision, date, and any supplements) for each of the TS referenced topical reports used to prepare the COLR.

5

Enclosure Description and Assessment of Proposed License Amendment 4.2. Precedent Precedent has been established for the implementation of Framatome fuel in Combustion Engineering pressurized water reactors (PWR). Most recently, St. Lucie Unit 2 was approved for the implementation of Framatome Combustion Engineering 16x16 HTP' fuel design with M5 as a fuel rod cladding material as documented in ADAMS accession number ML16063A121 (Reference 18). St. Lucie Unit 2 was granted an exemption from the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50 to allow the use of M5 fuel rod cladding as documented in ADAMS Accession Number ML16015A286 (Reference 18).

4.3. Regulatory Discussion of Topical Reports The proposed implementation of Framatome fuel requires changes to TS 2.1.1.2, TS 4.2.1 and TS 5.6.5.

The restrictions of TS 2.1.1.2 prevent overheating of the fuel and cladding and possible cladding perforation that would result in the release of fission products to the reactor coolant. The current TS 2.1.1.2 text is consistent with the Standard Technical Specifications for Combustion Engineering Plants as specified in NUREG-1432 (References 3 and 4). The proposed change will revise the PVNGS Units 1, 2, and 3 TS to clarify that the current TS 2.1.1.2 is applicable to Westinghouse supplied fuel. The proposed change will also revise the PVNGS Units 1, 2, and 3 TS to specify a peak fuel centerline temperature safety limit for Framatome supplied fuel that is consistent with the best estimate fuel temperature relationship, including uncertainty, defined by Equation 12-3 in approved COPERNIC Topical Report BAW-10231(P)(A) (Reference 10).

Technical Specification 4.2.1, Fuel Assemblies, provides a list of approved fuel rod cladding materials, including Zircaloy, ZIRLO and Optimized ZIRLO'. Optimized ZIRLO' was previously added to this list following approval of a permanent exemption request from the requirements of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, and 10 CFR 50 Appendix K, ECCS Evaluation Models.

The proposed change will revise the PVNGS Units 1, 2, and 3 TS to allow the use of M5 fuel rod cladding material. This change will be implemented by simplifying the TS 4.2.1 list of fuel rod cladding materials to the phrase zirconium-alloy clad. The proposed change will also revise the PVNGS Units 1, 2, and 3 TS to add a paragraph break prior to the sentence beginning A limited number of lead test assemblies. This paragraph break is appropriate because the use of lead test assemblies is not related to the TS 4.2.1 text characterizing the fuel assemblies. The proposed change will also revise the PVNGS Units 1, 2, and 3 TS to delete the TS 4.2.1 sentence requiring an approved exemption of use of other cladding material.

This sentence is unnecessary because it does not impose any requirement that is not already imposed by 10 CFR 50.46. In addition, the removal of this sentence is consistent with the NUREG-1432, Standard Technical Specifications, Combustion Engineering Plants (Reference 3).

TS 5.6.5.b lists the documents which describe the analytical methods used to determine the core operating limits. The proposed change will revise TS 5.6.5.b by updating the references to be consistent with the analytical methods that will be used to determine the core operating limits following Framatome fuel implementation. The proposed change will add the following topical 6

Enclosure Description and Assessment of Proposed License Amendment reports to the list of referenced core operating analytical methods to be consistent with the analytical methods that will be used to determine the core operating limits following Framatome fuel, critical heat flux (CHF) correlation, and gadolinia burnable absorber methodology implementation.

EMF-2103P-A, Realistic Large Break LOCA Methodology for Pressurized Water Reactors

EMF-2328(P)(A), PWR Small Break LOCA Evaluation Model, S-RELAP5 Based

BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code

BAW-10241(P)(A), BHTP DNB Correlation Applied with LYNXT

EPRI-NP-2511-CCM-A, VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores EMF-2103P-A (Realistic Large Break LOCA Methodology)

Topical Report EMF-2103P-A (Reference 7) describes the Framatome methodology developed for the realistic evaluation of a large break loss-of-coolant accident (LBLOCA) for PWRs with recirculation (U-tube) steam generators. Specifically, Westinghouse 3-and 4-loop designs; Combustion Engineering (CE) 2x4 designs; and AREVA 3- and 4-loop designs, all with fuel assembly lengths of 14 feet or less and emergency core cooling system (ECCS) injection to the cold legs, are covered.

APS has demonstrated that the conditions and limitations contained in the NRC SE for this topical report will be met, as described in Attachment 5.

EMF-2328(P)(A) (Small Break LOCA Methodology)

Topical Report EMF-2328(P)(A) (Reference 8) documents use of the S-RELAP5 thermal-hydraulic analysis computer code for analysis of the small break loss of coolant accident (SBLOCA) for Westinghouse and Combustion Engineering PWRs. The ANF-RELAP code was modified to bring it up to a standard that incorporates the thermal-hydraulic code RELAP5/MOD2, the fuel design code RODEX2, the ICECON containment model, and the hot rod model code TOODEE2, into a single system calculation. In so doing, the RELAP5/MOD2 code was modified to include selected models from the RELAP5/MOD3 code, improved neutronics, and models necessary to satisfy the requirements of 10 CFR Part 50, Appendix K.

It is noted that a collection of errata was released in January 2008 to incorporate corrections to the BETHSY testing facility assessment.

Topical Report EMF-2328(P)(A), Supplement 1(P)(A) (Reference 9) provides additional modeling information regarding the manner in which the SBLOCA evaluation model (EM) will treat the following eight areas:

Spectrum of break sizes

Core bypass flow paths in the reactor vessel

Reactivity feedback 7

Enclosure Description and Assessment of Proposed License Amendment

Delayed reactor coolant pump (RCP) trip

Maximum accumulator / safety injection tank (SIT) temperature

Loop seal clearing

Break in attached piping

Core nodalization Each issue is explained as to its treatment within the EM and the basis for that treatment, followed by a direct reference to any specific alteration of the treatment described in the main body of Revision 0 of the topical report. These changes are intended to improve the rigor and completeness of the original methodology, while also addressing and resolving several NRC staff issues raised regarding the Framatome small-break methodology over the preceding several years.

There are no conditions and limitations contained in the NRC SE for this topical report supplement.

BAW-10231P-A (COPERNIC Code)

Topical Report BAW-10231P-A (Reference 10) describes the COPERNIC (fuel rod design computer code) that performs the thermal/mechanical analyses necessary to accurately simulate the behavior of a fuel rod during its irradiation. The COPERNIC code is approved for UO2 licensing applications with advanced cladding material, M5, to a peak rod average burnup of 62 GWd/MTU.

APS has demonstrated that the conditions and limitations contained in the NRC SE for this topical report will be met, as described in Attachment 5.

BAW-10241(P)(A) (BHTP DNB Correlation)

Topical Report BAW-10241(P)(A) (Reference 11) describes the HTP' departure from nucleate boiling (DNB) correlation and its proposed use with the LYNXT computer code.

The LYNXT form of the HTP' correlation is referred to as the BHTP CHF correlation.

Topical Report BAW-10241(P)(A), Revision 1, addresses the extension of the range of applicability of the independent variables in the BHTP CHF correlation.

It is noted that errata 1P-001 was released in February 2016 to incorporate corrections to the bundle conditions (test conditions) listed for one of the tests identified in Table A.3 of the topical report.

APS has demonstrated that the conditions and limitations contained in the NRC SE for this topical report will be met, as described in Attachment 5.

EPRI-NP-2511-CCM-A (VIPRE-01)

Topical Report EPRI-NP-2511-CCM-A (References 13, 14 and 15), VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores, describes the Versatile Internals and Components Program for Reactors; EPRI (VIPRE) computer code to assist utilities in performing detailed thermal-hydraulic analyses of reactor cores. The mathematical modeling used in the code is discussed in Volume 1. Volume 2 is the user's manual and 8

Enclosure Description and Assessment of Proposed License Amendment the programmer's manual are contained in Volume 3. Volume 4 documents the experimental data comparisons, sensitivity studies and plant behavior simulations. Input guidelines and capabilities and limitations of the code are presented in Volume 5.

Two versions of VIPRE-01 were submitted to the NRC for review. VIPRE-01 MOD01 was submitted in 1985 and approved in 1986. VIPRE-01 MOD01 is essentially restricted to PWR use only and is considered an outdated and obsolete version.

VIPRE-01 MOD02 was submitted to the NRC in 1990 to address issues raised from the review of MOD01 and to improve some of the limitations on boiling water reactor (BWR) usage. The VIPRE-01 MOD02 SE was issued in 1993. Since there were no substantive modeling changes which would impact PWR calculations, VIPRE-01 MOD02 was approved for PWR applications subject to the same limitations given in the VIPRE-01 MOD01 SE. The MOD02 version inherited the limitations from the MOD01 SE.

APS has demonstrated that the conditions and limitations contained in the NRC SE for these topical reports will be met, as described in Attachment 5.

Assumptions used for accident initiators and/or safety analysis acceptance criteria are not altered by the addition of these topical reports.

Use of the referenced methodologies will support implementation of Framatome fuel. The ability to use either Westinghouse or Framatome fuel will ensure the stability of the fuel supply.

4.4. No Significant Hazards Consideration Determination Arizona Public Service (APS) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No.

The proposed changes establish a COPERNIC fuel rod design computer code peak fuel centerline temperature safety limit for Framatome HTP' fuel, allows for the use of M5 fuel rod cladding material by simplifying the TS 4.2.1 list of fuel rod cladding materials to the phrase zirconium-alloy clad, and updates the TS 5.6.5.b list of documents describing the core operating limits report (COLR) analytical methods to implement Framatome fuel, BHTP critical heat flux (CHF) correlation, gadolinia burnable absorber, and VIPRE-01 (Versatile lnternals and Component Program for Reactors) code methodology.

The requested Technical Specification (TS) changes do not involve any plant modifications that could affect system reliability, component performance, or the possibility of operator error.

There is a new time requirement for an existing operator action, but it has been demonstrated to be able to be performed successfully well within the time requirement. The requested TS changes do not affect any postulated accident precursors, do not affect any accident mitigation systems, and do not introduce any new accident initiation methods. The response of the Framatome fuel to postulated accidents has been analyzed using the proposed safety limit, fuel 9

Enclosure Description and Assessment of Proposed License Amendment design characteristics, and associated methodologies. These evaluation results show that the fuel response to postulated accidents is within applicable acceptance criteria.

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

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

Response: No.

The proposed changes establish a COPERNIC peak fuel centerline temperature safety limit for Framatome HTP' fuel, allows for the use of M5 fuel rod cladding material by simplifying the TS 4.2.1 list of fuel rod cladding materials to the phrase zirconium-alloy clad, and updates the TS 5.6.5.b list of documents describing the COLR analytical methods to implement Framatome fuel, BHTP CHF Correlation, gadolinia burnable absorber, and VIPRE-01 code methodology.

Physical changes associated with Framatome HTP' fuel (e.g., M5 cladding, fuel assembly spacer grids, gadolinia as a burnable absorber, MONOBLOC' construction, FUELGUARD' lower tie plate) do not introduce any new accident initiators and do not adversely affect the performance of any structure, system, or component previously credited for accident mitigation.

Use of Framatome fuel with M5 cladding in Palo Verde Nuclear Generating Station reactor cores is compatible with the plant design and does not introduce any new safety functions for plant structures, systems, or components. The fuel design performs within the fuel design limits.

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

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

Response: No.

The proposed changes establish a COPERNIC peak fuel centerline temperature safety limit for Framatome HTP' fuel, allows for the use of M5 fuel rod cladding material by simplifying the TS 4.2.1 list of fuel rod cladding materials to the phrase zirconium-alloy clad, and updates the TS 5.6.5.b list of documents describing the COLR analytical methods to implement Framatome fuel, BHTP CHF Correlation, gadolinia burnable absorber, and VIPRE-01 code methodology.

The existing TS safety limits for fuel supplied by Westinghouse are not being changed. The proposed COPERNIC peak fuel centerline temperature safety limit provides assurance that Framatome HTP' fuel fission product barriers will perform within applicable acceptance criteria for normal operation, anticipated operational occurrences, and postulated accidents. The methodology implementing the BHTP CHF correlation for Framatome HTP' fuel ensures that the applicable margin of safety is maintained (i.e., there is at least 95% probability at a 95%

confidence level that the hot fuel rod in the core will not experience departure from nucleate boiling (DNB)).

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

10

Enclosure Description and Assessment of Proposed License Amendment Conclusion APS concludes that operation of the facility in accordance with the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c),

and, accordingly, a finding of "no significant hazards consideration" is justified. Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

5. ENVIRONMENTAL EVALUATION The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),

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

6. PERMANENT EXEMPTION - 10 CFR 50.46 AND 10 CFR PART 50 APPENDIX K FOR M5 The proposed change will revise the PVNGS Units 1, 2, and 3 TS to allow the use of M5 fuel rod cladding material. Acceptable fuel rod cladding material is identified in PVNGS Units 1, 2, and 3 Technical Specification (TS) 4.2.1, Fuel Assemblies. The proposed change will add M5 fuel rod cladding material as an acceptable material by simplifying the TS 4.2.1 list of fuel rod cladding materials to the phrase zirconium-alloy clad. A permanent exemption from certain requirements of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, and 10 CFR 50, Appendix K, is required to support this change. By letter dated October 14, 2008, the NRC staff approved a temporary exemption from these requirements to support the PVNGS Units 1, 2 and 3 Framatome Lead Fuel Assembly (LFA) program (References 1 and 2). The requested permanent exemption will replace the approved temporary exemption.

Part 50.46(a)(l)(i) of Title 10 of the Code of Federal Regulations (10 CFR 50.46(a)(l)(i)) states in part:

"Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical Zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section. ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties 11

Enclosure Description and Assessment of Proposed License Amendment sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated."

10 CFR 50.46 continues with a delineation of specifications for peak cladding temperature, maximum hydrogen generation, coolable geometry, and long-term cooling. Since 10 CFR 50.46 specifically refers to fuel with Zircaloy or ZIRLO cladding and doesnt list M5 cladding, the use of M5 cladding requires a permanent exemption from this section of the regulations.

10 CFR 50, Appendix K, paragraph I.A.5, states in part:

"The rate of energy release, hydrogen generation, and cladding oxidation from the metal/water reaction shall be calculated using the Baker-Just equation."

The Baker-Just equation presumes the use of Zircaloy or ZIRLO cladding. The routine use of M5 cladding requires a permanent exemption from this section of the regulations.

Topical Report EMF-2328(P)(A) (Reference 8) documents analysis of the SBLOCA for Westinghouse and Combustion Engineering PWRs. The SBLOCA topical report acknowledges that the Baker-Just metal-water reaction model is used for oxidation during the transient. Per Topical Report BAW-10240(P)-A (Reference 12), Baker-Just is approved for use with M5.

Topical Report EMF-2103P-A (Reference 7) describes the Framatome methodology developed for the realistic evaluation of a LBLOCA for PWRs with recirculation (U-tube) steam generators.

The realistic LBLOCA topical report acknowledges that energy released through the transient oxidation of cladding is calculated using the Cathcart-Pawel correlation for oxide layer growth, and that the Cathcart-Pawel equation is applicable to all zirconium based cladding alloys currently used, including M5 cladding.

Pursuant to 10 CFR 50.12, Specific Exemptions, APS is requesting a permanent exemption from the requirements of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, and 10 CFR 50, Appendix K, ECCS Evaluation Models, for PVNGS Units 1, 2 and 3.

The permanent exemption will allow the use of fuel assemblies manufactured by Framatome with M5 alloy clad fuel rods, consistent with NRC-approved APS and Framatome design and analysis methodologies.

The use of M5 alloy cladding has been approved by the NRC for several US PWRs. Most recently, St. Lucie Unit 1 received approval for the use of M5 in 2014 (Reference 16), and St.

Lucie Unit 2 received approval for the use of M5 in 2016 (Reference 18).

10 CFR 50.12, Specific Exemption The standards set forth in 10 CFR 50.12 provide that the Commission may grant exemptions from the requirements of the regulations for reasons consistent with the following:

The exemption is authorized by law;

The exemption will not present an undue risk to the public health and safety;

The exemption is consistent with the common defense and security; and 12

Enclosure Description and Assessment of Proposed License Amendment

Special circumstances are present.

This exemption is authorized by law. This exemption results in changes to the operation of the plant by allowing the use of M5 as fuel rod cladding material in lieu of Zircaloy or ZIRLO.

As previously stated, 10 CFR 50.12 allows the NRC to grant exemptions from the requirements of 10 CFR Part 50. The NRC staff has previously determined that granting of this type of proposed exemption will not result in a violation of the Atomic Energy Act of 1954, as amended, or the Commission's regulations. Therefore, as required by 10 CFR 50.12 (a)(1), this requested exemption is "authorized by law."

The exemption will not present an undue risk to public health and safety. The M5 zirconium-alloy cladding has very low corrosion and hydrogen pickup rates; providing substantial margin for end of life corrosion and hydrogen limits. This material has been used extensively both in Europe and the United States for fuel rod cladding. The material has been generically reviewed and accepted by the NRC for use on CE fuel designs as addressed in Topical Report BAW-10240(P)-A (Reference 12). Reloads with M5 cladding have been provided in the United States since 2000 and on CE-14 designs since 2006.

The exemption is consistent with the common defense and security. The ability to use either Westinghouse or Framatome fuel will ensure security (i.e., assurance) of the PVNGS fuel supply by providing for multiple fuel vendors with reliable fuel designs and geographically diverse manufacturing facilities. This change in fuel material used in the plant has no relation to plant security issues. Therefore, the common defense and security are not impacted by this exemption request.

Special circumstances are present. As set forth in 10 CFR 50.12(a)(2)(ii), which states that special circumstances are present whenever:

"Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule..."

10 CFR 50.46 identifies acceptance criteria for ECCS system performance at nuclear power facilities. The effectiveness of the ECCS in PVNGS Units 1, 2 and 3 will not be affected by the use of M5 clad fuel assemblies. Due to the similarities in the material properties of the M5 alloy to Zircaloy or ZIRLO as identified in Topical Report BAW-10240(P)-A (Reference 12) it can be concluded that the ECCS effectiveness would not be adversely affected.

The intent of paragraph I.A.5 of Appendix K to 10 CFR 50 is to apply an equation for rates of energy release, hydrogen generation, and cladding oxidation from metal-water reaction that conservatively bounds all post-LOCA scenarios. The approved Framatome methodology for evaluating SBLOCA events uses the Baker-Just equation, which is approved for use with M5.

The approved Framatome methodology for evaluating realistic LBLOCA events uses the Cathcart-Pawel equation which is applicable to all zirconium based cladding alloys currently used, including M5 cladding.

The regulations of 10 CFR 50.46 and 10 CFR Part 50, Appendix K, make no provision for use of fuel rods clad in a material other than Zircaloy or ZIRLO. Since the chemical composition of the M5 alloy differs from the specifications for Zircaloy or ZIRLO, a plant-specific exemption is required to allow the use of the M5 alloy as a cladding material at PVNGS. The expected 13

Enclosure Description and Assessment of Proposed License Amendment performance of M5 clad material meets the intent of the regulations, as discussed in the M5 Topical Report (Reference 12). Therefore, application of these regulations in this circumstance would not serve the underlying purpose of the rule and is not necessary to achieve the underlying purpose of the rule, so special circumstances exist.

7. REFERENCES

1. Letter from D. C. Mims (APS) to NRC of March 08, 2008, Palo Verde Nuclear Generating Station (PVNGS) Unit 1; Docket No. STN 50-528; Request for Temporary Exemption from the Provisions of 10 CFR 50.46 and 10 CFR 50, Appendix K for Lead Fuel Assemblies, (ADAMS Accession No. ML080790524).

2. Letter from J. R. Hall (NRC) to R. K. Edington (APS) of October 14, 2008, Palo Verde Nuclear Generating Station, Unit 1 - Temporary Exemption from the Requirements of 10 CFR Part 50, Section 50.46 and Appendix K (TAC No. MD8330), (ADAMS Accession Nos. ML082730003 and ML082730006).

3. NUREG-1432, Revision 4.0, Volume 1, Standard Technical Specifications - Combustion Engineering Plants: Specifications, (ADAMS Accession No. ML12102A165).

4. NUREG-1432, Revision 4.0, Volume 2, Standard Technical Specifications - Combustion Engineering Plants: Bases, (ADAMS Accession No. ML12102A169).

5. SECY-16-0033, Draft Final Rule - Performance-Based Emergency Core Cooilng System Requirements and Related Fuel Cladding Acceptance Criteria (RIN 3150-AH42), March 16, 2016 (ADAMS Accession No. ML15238A933).

6. Technical Specification Task Force (TSTF) Traveler Number 363, Revision 0, Revise Topical Report References in ITS 5.6.5, COLR, (ADAMS Accession No. ML040630088).

7. EMF-2103P-A, Revision 3, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, June 2016.

8. EMF-2328(P)(A), Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2001.

9. EMF-2328(P)(A), Revision 0, Supplement 1(P)(A), PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, December 2016.

10. BAW-10231P-A, Revision 1, COPERNIC Fuel Rod Design Computer Code, January 2004.

11. BAW-10241(P)(A), Revision 1 as amended by Errata 1P-001 (February 2016), BHTP DNB Correlation Applied with LYNXT, July 2005.

12. BAW-10240(P)-A, Revision 0, Incorporation of M5 Properties in Framatome ANP Approved Methods, May 2004.

13. EPRI NP-2511-CCM-A, Mod 02, Revision 3, VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores, Volume 1 through 4 (Revision 4, February 2001), and Volume 5 (March 1988), Electric Power Research Institute.

14. Letter from C. E. Rossi (NRC) to J. A. Blaisdell (UGRA Executive Committee), Acceptance 14

Enclosure Description and Assessment of Proposed License Amendment for Referencing of Licensing Topical Report, EPRI NP-2511-CCM, "VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores, Volumes 1, 2, 3 and 4, May 1, 1986.

15. Letter from A. C. Thadani (NRC) to Y. Y. Yung (VIPRE-01 Maintenance Group),

Acceptance for Referencing of the Modified licensing Topical Report, EPRI NP-2511-CCM, Revision 3, VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores, (TAC No. M79498), October 30, 1993.

16. Letter, Perry H. Buckberg (NRC) to Mano Nazar (NEE), St. Lucie Plant, Unit No. 2 -

Issuance of Amendment Regarding Transitioning to Areva Fuel (CAC No. MF5495), April 19, 2016 (ADAMS Accession No. ML16063A121).

17. Letter, Lisa M. Regner (NRC) to Mano Nazar (NEE), St. Lucie Plant, Unit 1 - Issuance of Amendment Regarding the Use of M5 Alloy Fuel Cladding in Core Reload Applications (TAC No. MF1817), March 31, 2014 (ADAMS Accession No. ML14064A129).

18. Letter, Perry H. Buckberg (NRC) to Mano Nazar (NEE), St. Lucie Plant, Unit No. 2 -

Exemption from the Requirements of 10 CFR 50.46 and Appendix K to 10 CFR Part 50 to Allow the Use of M5 Fuel Rod Cladding (CAC No. MF5494), April 19, 2016 (ADAMS Accession No. ML16015A286).

19. Letters, L. Raynard Wharton (NRC) to Gregg R. Overbeck (APS), Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendements Re: Various Administrative Controls (TAC Nos. MB1668, MB1669, and MB1670), October 15, 2001 (ADAMS Accession No. ML012880473) and Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Correction of an Administrative Error Re: Amendments for Various Administrative Controls (TAC Nos. MB1668, MB1669, and MB1670), October 22, 2001 (ADAMS Accession No. ML012950413).

15

ATTACHMENT 1 Regulatory Commitment

Enclosure Attachment 1 ATTACHMENT 1 Regulatory Commitment The following table identifies a regulatory commitment in this document. Any other statements in this submittal represent intended or planned actions. They are provided for information purposes and are not considered to be regulatory commitments.

TYPE SCHEDULED COMPLETION REGULATORY COMMITMENT DATE continuing one-time compliance (if applicable)

APS commits to incorporating into the Time Critical X Upon Action Program the requirement to stop all Reactor implementation Coolant Pumps within 5 minutes following of Framatome pressurizer pressure dropping below the RCP NPSH HTP' Fuel limits during a Small Break LOCA event. This new time requirement for LOCA mitigation operator action is discussed in section 7.2 of Attachment 10.

1

ATTACHMENT 2 Technical Specifications Page Mark-ups Affected Pages: 2.0-1, 4.0-1 and 5.6-7

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, Departure from Nucleate Boiling Ratio (DNBR) shall be maintained at 1.34.

2.1.1.2 In MODES 1 and 2, 2.1.1.2.1 The peak fuel centerline temperature for Westinghouse supplied fuel shall be maintained < 5080°F (decreasing by 58°F per 10,000 MWD/MTU for burnup and adjusting for burnable poisons per CENPD-382-P-A).

2.1.1.2.2 The peak fuel centerline temperature for Framatome supplied fuel shall be maintained

< 4901°F (decreasing by 13.7°F per 10,000 MWD/MTU for burnup).

2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

PALO VERDE UNITS 1,2,3 2.0-1 AMENDMENT NO. 183

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Palo Verde Nuclear Generating Station is located in Maricopa County, Arizona, approximately 50 miles west of the Phoenix metropolitan area. The site is comprised of approximately 4,050 acres. Site elevations range from 890 feet above mean sea level at the southern boundary to 1,030 feet above mean sea level at the northern boundary. The minimum distance from a containment building to the exclusion area boundary is 871 meters.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 241 fuel assemblies. Each assembly shall consist of a matrix of zirconium-alloy clad Zircaloy or ZIRLO or Optimized ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases.

A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. Other cladding material may be used with an approved exemption.

4.2.2 Control Element Assemblies The reactor core shall contain 76 full strength and 13 part strength control element assemblies (CEAs).

The control section for the full strength CEAs shall be either boron carbide with Alloy 625 cladding, or a combination of silver-indium-cadmium and boron carbide with Alloy 625 cladding.

The control section for the part strength CEAs shall be solid Alloy 625 slugs with Alloy 625 cladding.

(continued)

PALO VERDE UNITS 1,2,3 4.0-1 AMENDMENT NO. 172, 205

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 Core Operating Limits Report (COLR) (continued)

20. CENPD-382-P-A, Methodology for Core Designs Containing Erbium Burnable Absorbers. [Methodology for Specifications 3.1.1, Shutdown Margin-Reactor Trip Breakers Open; 3.1.2, Shutdown Margin-Reactor Trip Breakers Closed; and 3.1.4, Moderator Temperature Coefficient.]

21. CEN-386-P-A, Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 16 x 16 PWR Fuel. [Methodology for Specifications 3.1.1, Shutdown Margin-Reactor Trip Breakers Open; 3.1.2,Shutdown Margin-Reactor Trip Breakers Closed; and 3.1.4, Moderator Temperature Coefficient.]

22. WCAP-16500-P-A, "CE 16x16 Next Generation Fuel Core Reference Report." [Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR]

23. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis." [Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR]

24. CENPD-387-P-A, "ABB Critical Heat Flux Correlations for PWR Fuel." [Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR]

25. WCAP-16523-P-A, "Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes." [Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR]

26. WCAP-16072-P-A, "Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs." [Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR]

27. EMF-2103-P-A, Realistic Large Break LOCA Methodology for Pressurized Water Reactors. [Methodology for Specification 3.2.1, Linear Heat Rate]

28. EMF-2328 (P) (A), PWR Small Break LOCA Evaluation Model, S-RELAP5 Based. [Methodology for Specification 3.2.1, Linear Heat Rate]

PALO VERDE UNITS 1,2,3 5.6-7 AMENDMENT NO. 178,

Reporting Requirements 5.6 5.6 Reporting Requirements

29. BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code. [Methodology for Specification 3.2.1, Linear Heat Rate]

30. BAW-10241 (P) (A), BHTP DNB correlation Applied with LYNXT. [Methodology for Specification 3.2.4, DNBR]

31. EPRI-NP-2511-CCM-A, VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores. [Methodology for Specification 3.2.4, DNBR]

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 PAM Report When a report is required by Condition B or F of LCO 3.3.10, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.

5.6.8 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG.

b. Active degradation mechanisms found.

c. Nondestructive examination techniques utilized for each degradation mechanism.

PALO VERDE UNITS 1,2,3 5.6-8 AMENDMENT NO. 178,

ATTACHMENT 3 Clean Technical Specifications Pages Affected Pages: 2.0-1, 4.0-1, and 5.6-7

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, Departure from Nucleate Boiling Ratio (DNBR) shall be maintained at 1.34.

2.1.1.2 In MODES 1 and 2, 2.1.1.2.1 The peak fuel centerline temperature for Westinghouse supplied fuel shall be maintained < 5080°F (decreasing by 58°F per 10,000 MWD/MTU for burnup and adjusting for burnable poisons per CENPD-382-P-A).

2.1.1.2.2 The peak fuel centerline temperature for Framatome supplied fuel shall be maintained

< 4901°F (decreasing by 13.7°F per 10,000 MWD/MTU for burnup).

2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

PALO VERDE UNITS 1,2,3 2.0-1 AMENDMENT NO. 183

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Palo Verde Nuclear Generating Station is located in Maricopa County, Arizona, approximately 50 miles west of the Phoenix metropolitan area. The site is comprised of approximately 4,050 acres. Site elevations range from 890 feet above mean sea level at the southern boundary to 1,030 feet above mean sea level at the northern boundary. The minimum distance from a containment building to the exclusion area boundary is 871 meters.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 241 fuel assemblies. Each assembly shall consist of a matrix of zirconium-alloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases.

A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Element Assemblies The reactor core shall contain 76 full strength and 13 part strength control element assemblies (CEAs).

The control section for the full strength CEAs shall be either boron carbide with Alloy 625 cladding, or a combination of silver-indium-cadmium and boron carbide with Alloy 625 cladding.

The control section for the part strength CEAs shall be solid Alloy 625 slugs with Alloy 625 cladding.

(continued)

PALO VERDE UNITS 1,2,3 4.0-1 AMENDMENT NO. 205