ML15336A087

Palo Verde, Units 1, 2, and 3 - License Amendment Request to Revise Technical Specifications to Incorporate Updated Criticality Safety Analysis

ML15336A087
Person / Time
Site:	Palo Verde
Issue date:	11/25/2015
From:	Lacal M L Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15336A251	List: ML15336A083 ML15336A084 ML15336A087
References
102-07149-MLL/TNW
Download: ML15336A087 (96)

Text

Enclosure Attachment 8 contains PROPRIETARY information to be withheld under 10 CFR 2.39010 CFR 50.90Maria L. LacalVice President, Nuclears~ale Regulatory

&OversightVed Nuclear Generating StationP.O. Box 52034Phoenix, AZ 85072/° Mail Station 7605Tel 623.393.6491 102-07149-M LL/TNWNovember 25, 2015U. S. Nuclear Regulatory Commission ATI-N: Document Control DeskWashington, DC 20555-0001

Subject:

Palo Verde Nuclear Generating Station (PVNGS)Units 1, 2, and 3Docket Nos. STN 50-528, 50-529, and 50-530License Amendment Request to Revise Technical Specifications toIncorporate Updated Criticality Safety AnalysisIn accordance with the provisions of Section 50.90 of Title 10 of the Code of FederalRegulations (10 CFR), Arizona Public Service Company (APS) is submitting a request for alicense amendment to revise the Technical Specifications (TS) for Palo Verde NuclearGenerating Station Units 1, 2, and 3. The proposed amendment would modify TSrequirements to incorporate the results of an updated criticality safety analysis for both newand spent fuel storage.The enclosure to this letter provides a description and assessment of the proposed changesincluding a technical evaluation, a regulatory evaluation, a significant hazards consideration, and an environmental consideration.

The enclosure also contains eight attachments.

Attachment 1 provides the marked-up existing TS pages. Attachment 2 provides the revised(clean) TS pages. Attachment 3 provides the marked-up TS Bases pages to show theproposed changes.This submittal contains new regulatory commitments (as defined by NEI 99-04, Guidelines for Managing NRC Commitment

Changes, Revision

0) to be implemented, which areidentified in Attachment

4. Attachment 5 provides a non-proprietary version of the criticality safety analysis.

Attachment 6 provides a material qualification report for NETCO-SNAP-IN neutron absorbing spent fuel pool rack inserts.Attachment 7 is an affidavit signed by Westinghouse Electric Company LLC that sets forththe basis on which the proprietary information in Attachment 8 may be withheld from publicdisclosure by the Commission and addresses with specificity the considerations listed in 10CFR 2.390(b)(4).

Correspondence with respect to the proprietary aspects of Attachment 8A member of the STAR!S (Strategic Teaming and Resource Sharing)

Alliance Callaway

• Diablo Canyon " Palo Verde

• Wolf CreekAttachment 8 transmitted herewith contains PROPRIETARY information.

When separated from Attachment 8, this transmittal document is decontrolled.

102-07149-M LL/TNWA-TEN: Document Control DeskU. S. Nuclear Regulatory Commission LAR to Incorporate Updated Criticality Safety Analysis in TSPage 2or the supporting Westinghouse affidavit should reference Westinghouse letter numberCAW-15-4271 and be addressed to James A. Gresham,

Manager, Regulatory Compliance, Westinghouse Electric

Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry

Township, Pennsylvania 16066.Attachment 8 is the Criticality Safety Analysis for Palo Verde Nuclear Generating StationUnits 1, 2, and 3, WCAP-18030-P (Proprietary),

which contains information proprietary toWestinghouse Electric Company LLC.A public pre-submittal meeting was held with the NRC on May 11, 2015 (Agency DocumentAccess and Management System [ADAMS] accession number ML15140A314) to discuss thecriticality safety analysis performed in support of this license amendment request.

A follow-up public conference call to address action items from the May 11, 2015, pre-submittal meeting was held on September 1, 2015 (ADAMS accession number ML15286A028).

In accordance with the PVNGS Quality Assurance

Program, the Plant Review Board and theOffsite Safety Review Committee have reviewed and approved the proposed amendment.

By copy of this letter, this license amendment request is being forwarded to the ArizonaRadiation Regulatory Agency in accordance with 10 CFR 50.91(b)(1).

APS requests approval of the proposed license amendment by October 1, 2017, and willimplement the TS amendment within 90 days following NRC approval.

This request isnecessary to complete the Spent Fuel Pool Transition Plan by the end of 2019.Should you have any questions concerning the content of this letter, please contact ThomasWeber, Department Leader, Nuclear Regulatory

Affairs, at (623) 393-5764.

I declare under penalty of perjury that the foregoing is true and correct.Executed on z 2-< .(Date)Sincerely, M LL/TN W/J R/af

Enclosure:

Description and Assessment of Proposed License Amendment cc: M. L. Dapas NRC Region IV Regional Administrator M. M. Watford NRC NRR Project Manager for PVNGSL. J. KIoss NRC NRR Project ManagerC. A. Peabody NRC Senior Resident Inspector for PVNGSA. V. Godwin Arizona Radiation Regulatory'Agency (ARRA)T. Morales Arizona Radiation Regulatory Agency (ARRA)

Enclosure Description and Assessment of Proposed License Amendment TABLE OF CONTENTS1.0 SUMMARY DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Proposed Changes to the Technical Specifications 2.2 Need for Proposed Changes3.0 TECHNICAL EVALUATION 3.1 Spent Fuel Pool Analysis3.2 New Fuel Storage and Fuel Transfer Equipment Analysis3.3 Spent Fuel Pool Transition Plan4.0 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements

4.2 Precedent

4.3 Significant Hazards Consideration

4.4 Conclusion

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

ATTACHMENTS

1. Marked-up Technical Specifications Pages2. Revised Technical Specifications Pages (Clean Copy)3. Marked-up Technical Specifications Bases Pages4. List of Regulatory Commitments

5. Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3(Non-proprietary),

WCAP-1 8030-NP, Revision 0, September 20156. Material Qualification Report of MAXUS for Spent Fuel Storage, NET-300047-07 Rev 1,November 20157. Westinghouse Application for Withholding Proprietary Information from PublicDisclosure, CAW-15-4271, September 3, 20158. Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3(Proprietary),

WCAP-18030-P, Revision 0, September 2015 LIST OF ACRONYMSANP AREVA PVNGS Lead Test Assembly Combustion Engineering 16x16 FuelAPS Arizona Public Service CompanyENDF Evaluated Nuclear Data FileFHE Fuel Handling Equipment IFSR Intermediate Fuel Storage RackLAR License Amendment RequestLER Licensee Event ReportNFS New Fuel StorageNGF Combustion Engineering 16x16 Next Generation FuelPVNGS Palo Verde Nuclear Generating StationSFP Spent Fuel PoolSTD Standard Combustion Engineering 16x16 FuelTS Technical Specification(s)

VAP Value Added Pellet Combustion Engineering 16x16 Fuelii Enclosure Description and Assessment of Proposed License Amendment 1.0 SUMMARY DESCRIPTION The proposed amendment would revise Palo Verde Nuclear Generating Station (PVNGS)Renewed Operating License Nos. NPF-41, NPF-51, and NPF-74 to amend the Technical Specifications (TS) to incorporate the results of the updated criticality safety analysis, WCAP-1 8030-P (Reference 6.1). The proposed amendment will correct a non-conservative TSregarding NRC approved License Amendment Number 125, which describes the currentlicensing bases for the criticality safety analysis for PVNGS. This is further discussed in Section2.2 of this proposed amendment.

This enclosure provides a description and assessment of the proposed changes including atechnical evaluation, a regulatory evaluation, a significant hazards consideration, and anenvironmental consideration.

The enclosure also contains eight attachments.

Attachment 1provides the marked-up existing TS pages. Attachment 2 provides the revised (clean) TSpages. Attachment 3 provides the marked-up TS Bases pages to show the proposed changes.This submittal contains new regulatory commitments (as defined by NEI 99-04, Guidelines forManaging NRC Commitment

Changes, Revision

0) to be implemented, which are identified inAttachment

4. Attachment 5 provides a non-proprietary version of the criticality safety analysis.

Attachment 6 provides a material qualification report for NETCO-SNAP-IN neutron absorbing spent fuel pool rack inserts.Attachment 7 is an affidavit signed by Westinghouse Electric Company LLC that sets forth thebasis on which the proprietary information in Attachment 8 may be withheld from publicdisclosure by the Commission and addresses with specificity the considerations listed in 10CFR 2.390(b)(4).

Attachment 8 is the Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units1, 2, and 3, WCAP-1 8030-P (Proprietary),

which contains information proprietary toWestinghouse Electric Company LLC.2.0 DETAILED DESCRIPTION 2.1 Proposed Changes to the Technical Specifications The following specific TS changes are proposed as part of the updated criticality safety analysis.

Marked-up TS pages are provided in Attachment 1 and revised (clean) TS pages are providedin Attachment 2.TS 3.7.17, Spent Fuel Assembly Storage-Revise the LCO statement and surveillance requirement statement to reflect theupdated spent fuel assembly storage requirements resulting from WCAP-18030-P

-Add Tables 3.7.17-1 through 3.7.17-5 to define the Fuel Regions-Replace Figure 3.7.17-1 to define the Allowable Storage Arrays-Delete Figures 3.7.17-2 and 3.7.17-3 (information is replaced by Tables 3.7.17-1through 3.7.17-5)

I Enclosure Description and Assessment of Proposed License Amendment

• TS 4.3.1, Criticality

-Change TS 4.3.1.1 .a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1.1.c to read "1460 ppm" [parts per million]

instead of "900 ppm"-Change TS 4.3.1.1.e to refer to Fuel Regions 1 -6as shown in Tables 3.7.17-1through 3.7.17-5-Delete TS 4.3.1.1.f through 4.3.1.1.h (Fuel Regions are defined in Tables 3.7.17-1through 3.7.17-5)

-Change TS 4.3.1 .2.a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1 .2.d to replace "A nominal 17 inch center to center..."

with"A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center..."

• Add new program TS 5.5.21, Spent Fuel Storage Rack Neutron Absorber Monitoring Program (Proposed TS 5.5.20, Risk In formed Completion Time Program, was submitted on July 31, 2015 [Agency Document Access and Management System (ADAMS)accession number ML15218A300])

The TS Bases will also be revised for consistency with the proposed TS changes and withWCAP-1 8030-P. A markup of the TS Bases pages reflecting these changes is provided inAttachment 3 for information.

The proposed TS Bases changes will be implemented inaccordance with TS 5.5.14, Technical Specifications (TS) Bases Control Program, at the sametime that the TS changes in the approved license amendment request (LAR) are implemented.

2.2 Need for Proposed ChangesIn March of 2000, the NRC approved License Amendment Number 125, which describes thecurrent licensing bases for the criticality safety analysis for PVNGS. That amendment increased the storage capacity of the spent fuel pools (SFPs) by allowing credit for soluble boron anddecay time in the criticality safety analysis.

The amendment also increased the maximumradially averaged fuel enrichment from 4.3 weight percent U-235 to 4.8 weight percent U-235.The methodology that was the basis for that amendment was analogous to that developed inWCAP-14416-P-A, Westinghouse Spent Fuel Rack Criticality Analysis Methodology, which wasreviewed and approved by the NRC for use [NRC Letter, T. E. Collins (NRC) to T. Greene(WOG), "Acceptance for Referencing of Licensing Topical Report WCAP-14416-P, Westinghouse Spent Fuel Rack Methodology (TAC No. M93254),"

dated October 25, 1996]. In2001 and 2004, Arizona Public Service Company (APS) submitted license amendment requests(LARs) to the NRC that would support replacement of the steam generators and authorize subsequent operation at an increased maximum power level of 3990 Megawatts thermal (a 2.94percent increase).

The NRC approved the amendments for Unit 2 in 2003 (ADAMS accession number ML032720538) and for Units I and 3 in 2005 (ADAMS accession numberML0531 30275).In May of 2013, APS submitted Licensee Event Report (LER) 2013-001-00 (ADAMS accession number ML13133A002),

which reported that certain impacts to the SFP criticality safety analysisapproved in License Amendment 125 had not been considered by APS during the increase inthe maximum power level to 3990 MWt. One of the corrective actions in the LER was to revisethe SFP criticality safety analysis using updated methodology and input parameters and to2 Enciosure Description and Assessment of Proposed License Amendment submit a LAR to correct the non-conservative TS. The criticality safety analysis methodology included in this LAR is based upon the most recent NRC approved guidance of Staff GuidanceRegarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, DSS-ISG-201 0-01(Reference 6.2) and satisfies the actions stipulated in the PVNGS Corrective Action Programand PVNGS LER 201 3-001-00.

3.0 TECHNICAL EVALUATION This LAR documents an updated criticality safety analysis for the PVNGS SFPs, new fuelstorage (NFS) racks, interim fuel storage rack (IFSR) and fuel handling equipment (FHE).Attachment 8 is the plant-specific Westinghouse WCAP-18030-P, Revision 0, Criticality SafetyAnalysis for Palo Verde Nuclear Generating Station Units 1, 2 and 3. A plant-specific NETCO-SNAP-IN0 material qualification report is also included (Attachment

6) since the LAR credits thepresence of neutron poisons in the NETCO-SNAP-IN neutron absorber inserts.

The main bodyof the LAR includes descriptions and summary evaluations, while the attached WCAP andNETCO report provide additional details on topics that include computer codes, fuel designhistory, depletion

analysis, criticality

analysis, as well as the interface, normal, and accidentconditions for PVNGS.The change to TS 4.3.1 .2.d regarding NFS rack spacing is proposed to more accurately reflectthe NFS as-built drawings and the existing NFS criticality safety analysis of record. The text ofTS 4.3.1.2 states "The new fuel storage racks are designed and shall be maintained with...",

which refers to physical dimensions.

Therefore, TS 4.3.1 .2'.d must accurately reflect the as-builtdimensions that must be maintained throughout the life of the plant. The racks have a nominal18-inch center-to-center pitch on the short axis (east-west) and a 31-inch center-to-center pitchon the long axis (north-south).

According to both the existing NFS and updated criticality safetyanalyses, this rack design maintains keff <0.95 during all normal and accident conditions.

3.1 Spent Fuel Pool AnalysisDesign ApproachThe existing SEP storage racks are evaluated for the placement of fuel within the storage arraysdescribed in the proposed TS changes.

Credit is taken for the negative reactivity associated with burnup and post-irradiation cooling time (decay time). Additionally, some SEP storagearrays credit the presence of the neutron poison in the NETCO-SNAP-IN inserts.

Finally, creditis taken for the presence of soluble boron in the SFP for specific conditions.

Compliance for the SFP is demonstrated by establishing limits on the minimum allowable burnup as a function of initial enrichment and decay time for each fuel storage array. Aconservative combination of best estimate and bounding values has been selected to model thefuel in the analysis to ensure that fuel represented by the proposed TS is less reactive than thefuel modeled in the analysis.

Therefore, burnup limits will conservatively bound fuel to be storedin the SFP.3 Enclosure Description and Assessment of Proposed License Amendment Acceptance CriteriaThe objective of the SEP criticality safety analysis is to ensure that the SEP operates within thebounds of 10 CFR 50.68(b)(4):

• If no credit for soluble boron is taken, the keff of the spent fuel storage racks loaded withfuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percentprobability, 95 percent confidence level, if flooded with unborated water.* If credit is taken for soluble boron, the keff of the spent fuel storage racks loaded with fuelof the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percentprobability, 95 percent confidence level, if flooded with borated water, and the keff mustremain below 1 .0 (subcritical),

at a 95 percent probability, 95 percent confidence level, ifflooded with unborated water.Computer CodesThe SEP criticality safety analysis methodology employs the following computer codes andcross-section libraries:

• SCALE 6.1.2 (Reference 6.3) with the 238-group cross-section library based onEvaluated Nuclear Data File (ENDF)/B-VII.

• The two-dimensional transport lattice code PARAGON (Reference 6.4) and its 70-groupcross-section library based on ENDF/B-VI.3 PARAGON is used for simulation of in-reactor fuel assembly depletion to generate isotopics forburnup credit applications in the SEP. PARAGON is generically approved for depletion calculations (Reference 6.4) and has been chosen for this spent fuel criticality safety analysisbecause it has all the attributes needed for burnup credit applications.

There are no SafetyEvaluation Report limitations for the use of PARAGON in UO2 criticality analyses.

Additional discussion of the computer codes is provided in Section 2.3 of Attachment 8.Code Validation ProcessThe validation of the ENDF/B-VII library with the SCALE 6.1.2 CSAS5 module is documented inAppendix A of Attachment

8. The code validation shows that SCALE 6.1.2 is an accurate toolfor calculation of kefffor the applications in this LAR. The benchmark calculations utilize thesame computer platform and cross-section libraries that are used for the design basiscalculations.

3.1.1 Fuel and Fuel Storage Descriptions Four fuel designs were considered for the criticality safety analysis:

• Standard Combustion Engineering 16x16 (STD)* Value Added Pellet Combustion Engineering 16x16 (VAP)* Next Generation Fuel Combustion Engineering 16x16 (NGF)* AREVA PVNGS Lead Test Assembly Combustion Engineering 16x16 (ANP)4 Enclosure Description and Assessment of Proposed License Amendment Section 3.1.2 of Attachment 8 discusses the non-mechanical fuel features which are important to criticality safety and how they impact the number of distinct fuel designs to be considered inthe analysis.

PVNGS has three SFPs, one in each unit, that are identical in layout. Each SFP contains asingle rack design and each SEP is surrounded by a concrete wall with a stainless steel liner.The presence of neither the SEP concrete wall nor liner is credited in the criticality safetyanalysis.

All storage arrays are conservatively assumed to be radially infinite.

The SEP storage racks are made up of individual

modules, each of which is an array of fuelstorage cells. The storage racks are comprised of 17 modules:

twelve 8x9 arrays, four 8x12arrays, and one 9x9 array. The storage racks are stainless steel honeycomb structures withrectangular fuel storage cells compatible with fuel assembly materials and the spent fuelborated water environment.

The fuel assembly spacing of a nominal 9.5 inches center-to-center distance between adjacent storage cell locations is a minimum value after allowances are madefor rack fabrication tolerances and predicted deflections resulting from a safe shutdownearthquake.

The storage racks are designed to maximize the number of storage cells available (minimize storage cell pitch) to be used in a checkerboard pattern of fresh (i.e., new) fuel and emptystorage locations.

Trashcans in the SEP store various non-fissile materials, such as discarded control element assemblies and in-core instrumentation tubes, filters, and reconstitution materials.

The SEP racks contain a stainless steel L-insert in every other cell location as shown in PVNGSUpdated Final Safety Analysis Report (UFSAR) Figure 9.1-5 to help center the new fuelassemblies within this space. The stainless steel L-inserts are offset 11/16 inch from the cellwall. This reduces the positional uncertainty or the eccentric loading positions of fuel in theracks. There is not currently a surveillance program for the stainless steel L-inserts and one willnot be included as part of this submittal for the following reasons:* The industry has a large body of operating experience with stainless steel in SEPenvironments in which nothing suggests the stainless steel L-inserts would becomeincapable of performing their function.

• The stainless steel L-inserts are analogous to the SEP racks in their function and there isno precedent for requiring a surveillance program of the racks themselves.

A NETCO-SNAP-IN rack insert will not be installed in a cell that contains a stainless steelL-insert.

The NETCO-SNAP-IN rack inserts are described in Attachment 6.3.1.2 Depletion AnalysisThe depletion analysis is a vital part of any SEP criticality safety analysis which uses burnupcredit. The isotopic inventory of the fuel as a function of burnup is generated through thedepletion

analysis, thus the inputs used need to be carefully considered.

Section 4 ofAttachment 8 describes the methods used to determine the appropriate inputs for thegeneration of isotopic number densities to conservatively bound fuel depletion and storage.Some of the salient points regarding the approach to the depletion analysis include:5 Enclosure Description and Assessment of Proposed License Amendment

• The isotopic number densities generated by the fuel depletion calculations aredifferentiated by fuel enrichment and decay time after discharge.

The fuel has isotopicnumber densities which are calculated at enrichments of 3.0, 4.0, and 5.0 weight percentU-235 and decay times of 0, 5, 10, 15, and 20 years.* The soluble boron concentration in the reactor during operation impacts the reactivity offuel being discharged to the SEP. Boron is a strong thermal neutron absorber and itspresence hardens the neutron energy spectrum in the core, creating more plutonium.

Itis important to account for the presence of soluble boron during reactor operation toensure this impact is adequately accounted for in the isotopic generation.

• The fuel temperature during operation impacts the reactivity of fuel being discharged tothe SFP. Increasing fuel temperature increases resonance absorption in U-238 due toDoppler broadening which leads to increased plutonium production, increasing thereactivity of the fuel.* The limiting distributed axial burnup profiles are used with the uniform axial burnupprofile to calculate the burnup limits.* The limiting axial moderator temperature profiles are used with axially distributed anduniform burnup profiles to calculate the isotopics used in generating the burnup limits.Selecting an appropriate moderator temperature profile is important as it impacts themoderator density and the neutron spectrum during depletion, as discussed inNUREG/CR-6665 (Reference 6.5). An appropriate moderator temperature ensures theimpact of moderator density on the neutron spectral effects is bounded, conservatively biasing the isotopic inventory of the fuel.* Burnable absorber usage has been considered for the analysis and conservative assumptions have been utilized to bound the effects of burnable absorbers on fuelisotopics.

The burnable absorbers that have been used include both discrete andintegral burnable absorbers.

• The PVNGS fuel management strategy uses radial enrichment zoning to control fuel rodpower peaking.

Individual assemblies may contain two or three different fuel rodenrichments which are used to control peaking factors.

A study was performed todetermine the reactivity impact of operating with radial enrichment zoning instead ofuniform radial zoning.* All four of the different fuel designs listed in Section 3.1.1 and the conditions in whichthose designs were operated, or are planned to be operated, were considered in thedepletion analysis.

It became clear that the NGF design would be limiting throughout life.Therefore, the NGF design was used to develop the isotopics used in the spent fuelreactivity calculations.

• The parameters used in the final depletion calculations include core operation parameters, fuel assembly dimensions, axial burnup profiles, and moderator temperature profiles.

3.1.3 Spent Fuel Pool Criticality AnalysisKENO is the criticality code used to determine the absolute reactivity of burned and fresh fuelassemblies loaded in storage arrays. The dimensions and tolerances of the design basis fuelassembly and the fuel storage racks are the basis for the KENO models used to determine theburnup requirements for each fuel storage array, and to confirm the safe operation of the SEP6 Enclosure Description and Assessment of Proposed License Amendment under normal and accident conditions.

The trashcan characteristics are also modeled in thecriticality analysis.

Differences between fuel types include changes in fuel rod dimensions, such as pellet andcladding dimensions, and structural components, such as grid material and volumes.

Each ofthe fuel types which have been, or are planned to be, operated at the plant were considered.

The bounding fuel assembly design for the analysis has been determined as described inSection 4.3 of Attachment 8.Burnup Limit Generation To ensure safe operation of the PVNGS SFPs, the analysis defines fuel storage arrays whichdictate where assemblies can be placed in the SFPs based on enrichment (weight percent U-235), average burnup (GWd/MTU),

and decay time (years) since discharge.

Each assembly inthe reactor core depletes under slightly different conditions and can have a different reactivity atthe same burnup. This is accounted for in the analysis by using a combination of depletion parameters that together produce a bounding isotopic inventory throughout life. Additionally, while fuel manufacturing is a very tightly controlled

process, assemblies are not identical.

Reactivity margin is added to the KENO reactivity calculations for the generation of burnup limitsto account for manufacturing deviations.

Assembly storage is controlled by defining allowable storage arrays. An array can only bepopulated by assemblies of the fuel region defined in the array definition or a lower reactivity fuel region. Fuel regions are defined by assembly burnup, initial enrichment, and decay time.Reactivity biases are known variations between the real and analyzed system, and theirreactivity impact is added directly to the calculated keff. Uncertainties are random dispersions around a nominal, measured quantity.

Their impact is added to the calculated keff as the squareroot of the sum of the squares of the uncertainties.

The following biases and uncertainties areaccounted for in the analysis.

A detailed discussion of biases and uncertainties is provided inSection 5.2.3 of Attachment 8.* Reactivity effect of manufacturing tolerances

• Burnup measurement uncertainty

• Depletion uncertainty

• Fission product and minor actinide worth bias* An operational uncertainty of 0.002 Ak is an additional conservativism which is added tothe conservatism inherent in the specific power histories from reactor operation

• Eccentric fuel assembly positioning

• Uncertainty in the predictive capability of SCALE 6.1.2 and the associated cross-section library* SFP temperature bias within the allowable operating range* Borated and unborated biases and uncertainties 7

Enclosure Description and Assessment of Proposed License Amendment Interface Modelingq Interfaces are the locations where there is a change in either the storage racks or the storagerequirements of the fuel in question.

At PVNGS, each SFP has a single storage rack design.Therefore, the only interfaces that exist are those between arrays within the single storage rackdesign and the only interface conditions that need to be addressed in the analysis are thosebetween different fuel storage arrays. Additional details are provided in Section 5.3 ofAttachment 8.Normal Conditions Considered in the Criticality Safety AnalysisThere are five major types of normal conditions beyond the storage of fuel assemblies that areaddressed in the criticality safety analysis.

Type 1 conditions involve placement of components in or near the intact fuel assemblies whilenormally stored in the storage racks. This also includes removal and reinsertion of thesecomponents into the fuel when stored in the rack positions using specifically designed tooling.Examples include control element assemblies and guide tube inserts, such as in-coreinstrumentation tubes. The calculation results show that any components designed to beinserted into an assembly may be stored in a fuel assembly guide tube in the SFP.Type 2 conditions involve evolutions where the fuel assembly is removed from the normalstorage rack location for a specific procedure and returned to an allowable cell after completion of the procedure, such as fuel assembly

cleaning, inspection, reconstitution, or sipping.

Theseare bounded by the criticality analysis.

Fuel assembly reconstitution is a normal condition defined as either pulling damaged fuel pinsout of an assembly and reinserting intact pins with less reactivity than the damaged pin, or asremoving undamaged pins from a damaged assembly for insertion in a new assembly.

Damaged pins will be replaced with stainless steel pins or natural uranium pins. Additional information is provided in Section 5.4.2 of Attachment 8.Type 3 conditions involve inserting components that are not intact fuel assemblies into the fuelstorage rack cells. Examples include failed fuel rod baskets and miscellaneous maintenance equipment.

Any components that do not contain fissile materials can replace a fuel assembly ofany fuel region in one of the approved storage configurations.

Type 4 conditions include temporary installation of non-fissile components on the rackperiphery.

Analyses of the storage arrays contained in the criticality analysis assume an infinitearray of storage cells. This assumption bounds the installation of any non-fissile components onthe periphery of racks.Type 5 conditions involve miscellaneous conditions that do not fit into the first four normalcondition types. Examples include usage of fuel handling tools for their intended purpose,miscellaneous debris under the storage racks, and damaged storage cells.Section 5.4 of Attachment 8 provides further details about normal conditions within the SEP.8 Enclosure Description and Assessment of Proposed License Amendment Soluble Boron CreditIn accordance with 10 CFR 50.68, the criticality safety analysis ensures that the maximumcalculated ke., including all biases and uncertainties, meet the kef limit of less than 1.0(subcritical) if flooded with unborated water at a 95 percent probability, 95 percent confidence level. Additionally, the criticality safety analysis demonstrates that if the SFP is flooded withborated water, keff does not exceed 0.95, at a 95 percent probability, 95 percent confidence level.The minimum soluble boron concentration in the SEP to maintain keff < 0.95 for the limitingnormal condition including biases, uncertainties, and administrative margin is 450 ppm. Duringnormal operation, TS 3.7.15 requires a soluble boron concentration of> 2150 ppm, but the SFPboron concentration is maintained between 4000 and 4400 ppm in accordance with Technical Requirements Manual T3. 1.104, Borated Sources -Shutdown, and T3. 1.105, Borated Sources-Operating.

Consideration of Criticality Accidents in the SEPThe following reactivity-increasing accidents are considered and the analysis results areprovided in Section 5.6 of Attachment 8.SAssembly misload into the storage racks -this is the limiting accident which addresses both multiple assemblies being misloaded in series into unacceptable storage locations and the misload of a single assembly into an unacceptable storage location.

A multipleassembly misload is a hypothetical accident where assemblies are misloaded in seriesdue to a common cause. A single assembly misload requires 1100 ppm of boron tomaintain keff -< 0.95. A multiple assembly misload requires 1460 ppm of boron to maintainkef<0.95.* Spent fuel temperature outside operating range -the SEP is to be operated between60°F and 180°F, but under accident conditions this temperature could be higher.* Dropped and misplaced fresh assembly

-the analysis considers the dropping of the fuelassembly from the fuel handling machine during placement of the fuel assemblies in theracks. The dropped assembly could land horizontally on top of the other fuel assemblies in the rack. Additionally, the analysis considers the possibility to misplace a fuelassembly in a location not intended for fuel.* Seismic event -the SEP racks are seismic category I, designed and built to withstand the maximum potential earthquake stresses in this geographic area. Section 5.6.4 ofAttachment 8 provides additional details.o The spent fuel pool racks were originally designed to contain a 188 lb. neutronpoison insert in every cell. These neutron poison inserts were never installed.

Asthe mass of the original design is greater than the mass of the NETCO-SNAP-IN, the original analysis bounds the proposed change.* Inadvertent removal of a NETCO-SNAP-IN rack insert -this is a potential reactivity-increasing accident added by the incorporation of NETCO-SNAP-IN rack inserts.

Theabsence of an insert will cause a reactivity increase due to the loss of neutron absorbing material from the storage array.9 Enclosure Description and Assessment of Proposed License Amendment Fuel used to date at PVNGS has an initial radially averaged enrichment of < 4.55 weightpercent.

Limiting the maximum radially averaged enrichment to 4.65 weight percent mitigates the consequences of a multiple fuel assembly misload event without impacting operational flexibility.

There is no source of water within the fuel building that could reduce the boron concentration ofthe spent fuel pool from the value of 2150 ppm (Technical Specification LCO 3.7.15) to 1460ppm. A fire in the fuel building at elevation 140-ft. is the limiting event for boron dilution and itbounds all normal, seismic, and pipe break scenarios.

The current boron dilution analysisdemonstrates that the limiting boron dilution event, which is fighting a hypothetical fire on the140-ft level of the fuel building, will reduce the boron concentration from the TS limit of 2150ppm to 1900 ppm. This leaves adequate margin to the 1460 ppm credited by the SEP criticality safety analysis.

3.1.4 NETCO-SNAP-IN Rack InsertsThis proposed change would credit NETCO-SNAP-IN rack inserts for criticality control inindividual SEP storage rack cells to ensure that the requirements of TS 3.7.17 and theassociated WCAP-1 8030-P are maintained.

The NETCO-SNAP-IN rack inserts are credited inboth the borated and unborated conditions.

The installation of the NETCO-SNAP-IN rackinserts will be controlled as a design change implemented under the provisions of 10 CFR50.59, Changes, Tests and Experiments, from a structural,

seismic, and thermal-hydraulic perspective.

Attachment 6 describes the NETCO-SNAP-IN rack inserts, including their manufacture, anengineering evaluation, and corrosion testing information.

Neutron Absorber Monitoring Program (TS 5.5.21)Arizona Public Service will institute a performance-based long-term surveillance program for theNETCO-SNAP-IN inserts based on manufacturer recommendations, current industry operating experience, NEI guidance, and NRC safety evaluations for other plants that are using neutronabsorbing inserts.

The long-term surveillance program will evolve as information from thesesources changes and as the data from the PVNGS-specific inspections accumulate.

Thesurveillance program consists of periodic inspections of MAXUS material coupons fromsurveillance assemblies located in the SEPs and periodic inspection of full length inserts.Coupon Inspections Coupons will be selected from MAXUS production

material, identical to the material used tomanufacture the inserts, for periodic inspection.

Individual coupons will be subjected to pre-testand post-test characterizations.

As appropriate for each coupon type, coupon characterizations may include visual inspection, high resolution photography, neutron attenuation, stressrelaxation, blister and pit characterizations, as well as measurement of thickness, length, width,dry weight, and density.A surveillance assembly to which surveillance coupons are attached, also referred to as acoupon tree, will be placed in the related SEP prior to the first installation campaign of NETCO-SNAP-IN inserts and will reside there to support the monitoring program.

Periodically, coupons10 Enclosure Description and Assessment of Proposed License Amendment will be removed and sent to a qualified laboratory for testing.

The coupon trees in the relatedSFP will contain 48 general coupons, 24 galvanic couple coupons, and 24 bend coupons asdescribed below. They will be situated near the center of the active fuel region to maximizeexposure from the surrounding fuel. These coupons will be monitored for changes to theirphysical properties and for changes to their effective areal density or signs of corrosion, whichcould indicate neutron absorber material degradation.

The frequency for coupon removal and inspection is shown in Table 1.Table I -Frequency for Coupon Removal from RacksAfter 10 Years withCoupon Type First Ten Years Acpal efracGeneral 2 coupons every 2 years 2 coupons every 4 yearsBend I coupon every 2 years 1 coupon every 4 yearsGalvanic couples -304L stainless I couple every 6 yearsZircaloy 1 couple every 6 yearsInconel 718 1 couple every 6 yearsGeneral and Galvanic CouponsThe general coupons in each SFP are designed to carry the largest number of performance indicators for the insert material.

They will be subject to pre-examination, post-examination, andacceptance testing in accordance with Table 2.The galvanic couple coupons are composed of a MAXUS material coupon placed in contactwith Zircaloy, Inconel 718, or 304 stainless steel. Eight of each type will be used to produce atotal of 24 galvanic couples per coupon tree assembly in each SEP that will be subject to theinspections listed in Table 2.Bend CouponsOnce installed, the NETCO-SNAP-IN rack inserts assume a constant strain condition within theSEP storage rack cell. This compression leads to internal

stresses, especially at the bend, thatmight make the rack inserts susceptible to stress corrosion cracking.

An examination of theliterature on the subject indicates in general, that high-purity aluminum and low-strength aluminum alloys are not susceptible to stress corrosion cracking.

However, the surveillance bend coupons placed in the related SEP will be maintained under the same strain conditions asthe inserts to provide an indication of unexpected crack phenomena.

These coupons will beheld in capsules that compress them from their initial manufactured bend angle to an angle ofapproximately 90 degrees.

Table 3 provides the inspection details.11 Enclosure Description and Assessment of Proposed License Amendment Over time, the MAXUS material is expected to release some of the strain built up during theinstallation process.

The material has a metal matrix core made from 1000 series aluminum andboron carbide powder with an outer clad made from 5052 aluminum.

Existing literature for 1100series aluminum shows a stress relaxation rate of 58 percent over a period of 20 years. Giventhat 1100 series aluminum is a softer metal than 5052 aluminum, this rate is considered conservative for the MAXUS material due to the 5052 cladding.

The acceptance criterion forstress relaxation is 60 percent over a 20-year period. This rate will be used when determining minimum retention force requirements for the inserts during installation that will still hold theinserts in place during a seismic event after relaxation has occurred.

The bend coupon capsules will be removed from the coupon tree and sent to a qualified laboratory for testing where the coupons will be removed, thus relieving the strain on thecoupons and allowing them to return to an angle greater than 90 degrees.

The change ininternal stress can be correlated to the change in bend angle the coupon forms once it isremoved from the strained condition.

Deviation from the pre-characterized value will determine the amount of stress relaxation over the life of the coupon.The stress relaxation rate is not linear, rather it tends to follow a logarithmic pattern.

Therefore, a more significant loss of stress is expected in the first few years of exposure, but the relaxation rate becomes asymptotic over a longer period of time.*Table 2 -General and Galvanic Coupon Characterizations TetPre- Post- Acceptance I Rejection Characterization Characterization CriteriaVisual (high / Evidence of visualresolution digital indications ofphoto) performance inhibitors.

Dimension

/Min. thickness:

0.005 inch less thannominal thickness (excluding pit locations).

Thickness change: anychange of +0.010 inch I-0.004 inch (excluding pitlocations).

• Length change: anychange of +/- 0.02 inch* Width change: anychange of +/- 0.02 inchDensity /Any change of +/- 5%Areal density g/cm2 Boron-10minimum loading12 Enclosure Description and Assessment of Proposed License Amendment Weight loss as determined by dry change of +/- 5%weightCorrosion rate /< 0.05 mil/yrAnomaly characterization

/** To be determined at thetime of analysis.

• Length and width changes are not applicable for galvanic coupons.** At the presence of anomalies Table 3 -Bend Coupon Characterizations TetPre- Post- Acceptance

/Characterization Characterization Rejection CriteriaVisual (high Evidence of visualresolution indications ofdigital photo) performance inhibitors.

Thickness

' V Min. thickness:

0.005 inch less thannominal thickness (excluding pit locations).

Thickness change: anychange of +0.010 inch /-0.004 inch (excluding pit locations).

Bending V /Change in stress greaterstress than a rate of 60% over20 years *Weight loss as determined by dry change of +/- 5%weightAnomaly characterization

• • To be determined at thetime of analysis.

Stress relaxation rate is not linear. Stress relaxation will be re-evaluated if 60% is exceeded.

• • At the presence of anomalies Full-Length Insert Inspections The combined effects of adequate clearance and infrequent fuel assembly movement willpreclude significant wear of the rack insert. However, to verify NETCO-SNAP-IN materialperformance, a portion of the installed inserts will be subject to in-situ visual inspection andremoval for detailed inspection of wear performance.

For in-situ inspections at the frequency described in Table 4, rack inserts will be visually inspected by camera (while remaining in thestorage racks) to monitor for physical deformities such as bubbling, blistering, corrosion pitting,13 Enclosure Description and Assessment of Proposed License Amendment

cracking, or flaking.

Special attention shall be paid to the development of edge or cornerdefects.A region of high duty spent fuel storage rack cell locations shall be identified for full insertremoval and inspection.

These locations will be monitored for fuel insertion and removal eventsto ensure that their service bounds that of the general population of storage locations.

Onceevery 10 years, an insert will be fully removed from this region and will be inspected inaccordance with Table 5. The thickness measurements at several locations along the full insertlength will be compared with the as-built thickness measurements of the removed insert to verifyit has sustained uniform wear over its service life. A visual inspection of the removed insert willalso be performed.

Table 4 -Frequency for Full Insert Inspections Inspection Type First Ten Years Acpal efracIn-situ 2 inserts every 2 years 2 inserts every 4 yearsRemoval 1 insert every 10 yearsTable 5 -Full Insert Removal Inspection Characterizations TetPre- Post- Acceptance ICharacterization Characterization Rejection CriteriaVisual (high of visualresolution indications ofdigital photo) performance inhibitors.

Thickness thickness:

0.005 inch less thannominal thickness (excluding pitlocations).

Thickness change: anychange of +0.010 inch I-0.004 inch (excluding pit locations).

Retention force Retention force lessthan 50 lbs14 Enclosure Description and Assessment of Proposed License Amendment 3.1.5 Spent Fuel Pool Configuration Control (Human Performance Enhancements)

APS has a multi-tier defense-in-depth program to prevent and mitigate the severity of a scenarioin which multiple fuel assemblies are located in the wrong storage locations.

Specific aspects ofthis program are described below.Control of Move Sheet Generation

• Detailed administrative procedures for the generation of move sheets and revision ofmove sheets* Training and qualification of individuals responsible for generation and revision of movesheets for use and implementation of the new TS proposed in this LAR.* Graphical representation of approved arrays in TS 3.7.17 to minimize the probability ofmisinterpretations

• In accordance with the plant special nuclear materials procedures, APS maintains aRegion Specification Document for each SEP to aid the move sheet preparer and verifierin selecting and verifying proper placement of fuel assemblies.

This document tracks thefollowing for every fuel assembly at PVNGS:-Fuel assembly initial enrichment

-Fuel assembly burnup-Limiting Fuel Region in which the fuel assembly can be stored* Every fuel move is checked against the availability of the space and the eligibility of fuelto be stored there* A member of management confirms that each move sheet was generated in accordance with PVNGS procedures.

There are at least three signatures on each move sheet sent tothe field, including the move sheet preparer, the verifier, and management.

• A move sheet package is a change document that is used to specify and record changesto plant configuration as it relates to special nuclear material.

A move sheet contains asa minimum, the item to be moved, the "from" location, and the "to" location.

Any numberof fuel assemblies can be moved using a single move sheet package.Control of Fuel Movement* The spent fuel handling machine is only operated using approved procedures

• All individuals operating the spent fuel handling machine and acting as independent verifiers are trained in their position, including training on industry operating experience pertaining to fuel misload events* Fuel is moved only as directed by approved move sheets* The correct location of the spent fuel handling machine is independently verified before afuel move begins* The correct location of the spent fuel handing machine is independently verified beforethe fuel is placed in the SEP racks* Continuous communication is maintained between the fuel mover and verifier15 Enclosure Description and Assessment of Proposed License Amendment

• The fuel handlers visually confirm that fresh fuel is not placed in locations that are "face-adjacent" to other fresh fuel assemblies Use of Blockingq DevicesThe assumed limiting misload event at PVNGS involves placing a fuel assembly in a locationthat is required to be empty per TS 3.7.17. APS uses blocking devices to minimize theprobability of a fuel assembly being placed in one of these limiting locations.

Each blockintg device meets the following criteria:

• Physically configured to prevent insertion of a fuel assembly in a fuel storage location* Requires special tools to install or remove the blocking device from a storage location* The tool used to grapple a fuel assembly is physically incapable of grappling a blockingdevice* Designed to preclude falling into a storage location or becoming dislodged during normaloperation

• Will support the full load of a fuel assembly and the fuel assembly grappling tool* Allows continuous water flow through the storage cell0SIs easy to distinguish visually from a fuel assemblyBlocking devices are administratively controlled with the same level of rigor as fuelassemblies

• A blocking device move sheet package is a change document used to specify andrecord changes to plant configuration as it relates to blocking devices.

A blocking devicemove sheet contains, as a minimum, the "from" location and the "to" location of theblocking device. Any number of blocking devices may be moved using a single blockingdevice move sheet package.

Blocking devices will not be moved using the same movesheet package as fuel assemblies.

This restriction prevents a single error from removinga blocking device and placing a fuel assembly in a location that is required to be empty.Confirmation of Configquration ControlIn accordance with existing procedures, a 100 percent serial number check is performed onceper calendar year to ensure that every fuel assembly stored in the SFP matches the SFP maps.This provision limits the amount of time that a misload condition could potentially exist.Mitigqation of a Misload EventIf the controls discussed above are insufficient to prevent a fuel assembly from being misloaded, the following will mitigate the consequences of such an event:* Misload events, including misload events involving multiple fuel assemblies, have beenanalyzed-An adequate soluble boron margin mitigates the misload event. TS 3.7.15requires 2150 ppm of soluble boron* The limiting misload of a single fuel assembly requires 1100 ppm ofsoluble boron to maintain keff <0.9516 Enclosure Description and Assessment of Proposed License Amendment

• The limiting,

credible, multiple misload of placing a fresh fuel assemblyinto every blocked location in the most limiting array (Array C) requires1460 ppm of soluble boron to maintain keff <0.95Placing fresh fuel assemblies face-adjacent to one another is not credible.

A defense-in-depth approach provides

multiple, independent barriers to this event. These barriersinclude:-Move sheets are generated, independently

verified, and approved by qualified individuals

-Blocking devices or trash cans are placed in locations that are face-adjacent tolocations approved for the storage of fresh fuel-The fuel movers will verify that fresh fuel is not placed in face-adjacent locations prior to completing each fuel moveSufficient rigor is placed into the generation of move sheets, execution of fuel movement, andmaintenance of SEP maps that the likelihood of an assembly being misplaced in the SEP issmall. The misplacement of multiple fuel assemblies is less probable.

Therefore, the approachused in the analysis is appropriately conservative.

The impact of placing multiple fresh fuelassemblies in face-adjacent locations is not evaluated in the analysis because this event is notconsidered credible.

A regulatory commitment regarding the implementation of procedural controls to requireverification that fresh fuel assemblies are not placed face-adjacent to one another beforecompleting a fuel move is provided in Attachment 4.3.2 New Fuel Storage and Fuel Transfer Equipment AnalysisA criticality safety analysis was performed to support operation of the NFS racks, the IFSR, thenew fuel elevator, and the fuel upender and transfer machine.

When discussing the new fuelelevator, and the fuel upender and transfer machine together, they are referred to as the fuelhandling equipment (FHE). The existing NFS racks, IFSR, and FHE were evaluated to confirmthat each system maintains subcriticality while performing its designed purposes.

3.2.1 Storage and Equipment Description N.ew Fuel Storagqe DesigqnThe NFS rack assemblies are made up of individual racks similar to those shown in UFSARFigure 9.1-1. A minimum edge-to-edge spacing between fuel assemblies is maintained inadjacent rows. This spacing is the minimum value after allowances are made for rack fabrication tolerances and the predicted deflections resulting from postulated accident conditions.

The stainless steel construction of the storage racks is compatible with the water and thezirconium-clad fuel. The top structure of the racks is designed such that there is no openingbetween adjacent fuel cavities that is as large as the cross-section of the fuel bundle. Inaddition, the outer structure of the racks precludes the inadvertent placement of a bundleagainst the rack closer than the prescribed edge-to-edge spacing.17 Enclosure Description and Assessment of Proposed License Amendment Two concrete storage cavities are utilized for NFS. Each cavity is approximately 8 feet by23 feet and contains 45 fuel assemblies in stainless steel racks. Three racks are installed ineach cavity, forming a 3x15 array of fuel assemblies.

The rack structure provides at least 10 inches between the top of the active fuel and the top ofthe rack to preclude criticality in case a fuel assembly is dropped into a horizontal position onthe top of the rack. The NFS racks and facilities are qualified as Seismic Category I and willsurvive a safe shutdown earthquake without loss of safety function.

The following postulated accidents were considered in the design of the NFS racks:* Flooding

-complete immersion of the entire storage array in pure, unborated, roomtemperature water* Envelopment of the entire array in a uniform density aqueous foam or mist of optimumdensity that maximizes the reactivity of the finite array (a condition that could result fromfirefighting)

• A fuel assembly dropped from a height of 4.5 ft onto the rack that falls horizontally across the top of the rack* Tensile load of 5000 lbs on the rackAlthough the above accident conditions have been postulated, the FHE, NFS racks, and thebuilding arrangement are designed to minimize the possibility of these accidents and the effectsresulting from these accidents.

Intermediate Fuel Storagqe RackThe IFSR is a four-cavity fuel storage rack in a lx4 array designed as an intermediate storagelocation for fuel bundles during refueling.

The rack is located in the containment adjacent to thecore support barrel laydown area, which provides access to the refueling machine for insertion and removal of fuel bundles.Each cavity in the IFSR is a stainless steel can 8.69 inches on a side. The cavities areseparated by a fuel center-to-center pitch of 18.56 inches. Each of the cavities is open at thebottom to provide thermal cooling for the worst case fuel bundle. The rack structure is designedto maintain keff < 0.95 by assuring under all normal and accident conditions, which includesSSE, that the minimum edge distance is not violated and that a fuel bundle cannot violate the12-inch minimum stand-off distance around the cavities.

New fuel may be stored in the IFSR before being moved into the core. Partially spent fuel maybe moved out of the core and stored temporarily in the IFSR to provide spaces for fuel shuffling.

Spent fuel may be stored in the IFSR before being sent to the SFP.Fuel Upender and Transfer MachineThe transfer

machine, or carriage, conveys the fuel assemblies through the transfer tube. Twofuel assembly cavities are provided in the fuel carriage to reduce overall fuel handling time. Afterthe refueling machine deposits a spent fuel bundle in the open cavity, it only has to moveapproximately one foot to pick up the new fuel assembly, which was brought from the fuelbuilding in the other cavity. The handling operation in the fuel building is similar.

The dual cavity18 Enclosure Description and Assessment of Proposed License Amendment arrangement permits both fuel handling machines to travel fully loaded at all times. Fuelassemblies are placed on the transfer carriage in a vertical

position, lowered to the horizontal

position, moved through the fuel transfer tube on the transfer

carriage, and then restored to thevertical position.

Wheels support the carriage and allow it to roll on tracks within the transfer*tube. The track sections at both ends of the transfer tube are mounted on the upendingmachines to permit the carriage to be properly positioned at the limits of its travel.An upending machine is provided at each end of the transfer tube. Each machine consists of aStructural support base from which is pivoted an upending straddle frame that engages the two-cavity fuel carrier.

Hydraulic cylinders attached to the upending frame that engages the supportbase rotate the fuel carrier between the vertical and horizontal position.

A third fuel assemblywas modeled five inches from the transfer carriage to allow for the presence of an additional fuelassembly no closer than five inches from the carriage.

(Appendix B of Attachment 8)New Fuel ElevatorThe new fuel elevator is utilized to lower new fuel from the operating floor to the bottom of thepool where it is grappled by the spent fuel handling tool. The elevator is powered by a cablewinch and fuel is contained in a simple support structure whose wheels are captured in tworails.3.2.2 New Fuel Criticality Safety AnalysisAcceptance CriteriaThe objective of the criticality safety analysis is to ensure that the fuel storage operations arewithin the bounds 10 CFR 50.68(b)(2) and 50.68(b)(3):

• The estimated ratio of neutron production to neutron absorption and leakage (keff) of thefresh fuel in the storage racks shall be calculated assuming the racks are loaded withfuel of the maximum fuel assembly reactivity and flooded with unborated water and mustnot exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent suchflooding or if fresh fuel storage racks are not used.* If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racksare assumed to be loaded with fuel of the maximum fuel assembly reactivity and filledwith low-density hydrogenous fluid, the keff corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. Thisevaluation need not be performed if administrative controls and/or design featuresprevent such moderation or if fresh fuel storage racks are not used.Design ApproachCompliance is shown for the NFS racks, IFSR, and EHE by demonstrating that the system does not exceed 0.95 at a 95 percent probability with a 95 percent confidence level. Aconservative combination of best estimate and bounding values has been selected to model thefuel in the analysis to ensure that fuel represented by the proposed TS is less reactive than thefuel modeled in the analysis.

19 Enclosure Description and Assessment of Proposed License Amendment Computer CodesThe analysis methodology employs SCALE 6.1.2 (Reference 6.3) with the 238-group cross-section library based on ENDF/B-VII.

All analyses performed used the "Fresh fuel withoutAbsorber' validation suite. KENO-Va is used to determine the absolute reactivity of fresh fuelassemblies in the NFS.The validation of the ENDF/B-VII library with the SCALE 6.1.2 CSAS5 module is documented inAppendix A of Attachment

8. The code validation shows that SCALE 6.1.2 is an accurate toolfor calculation of keff for the applications in this LAR. The benchmark calculations utilize thesame computer plafform and cross-section libraries that are used for the design basiscalculations.

Limitingq Fuel Desigqn Selection There are four potentially limiting fuel designs that have been used at PVNGS. The VAP designis currently in use on site and NGF is planned for use in the future. Therefore, both VAP andNGF designs are considered the two potentially limiting fuel designs.

The STD and ANP fueldesigns do not need to be addressed for these calculations because they are bounded by theVAP fuel design. Additional information is provided in Section B.4.2 of Attachment 8.Treatment of ConcreteConcrete is a material which has a large variety of different potential compositions, all of whichcan be labeled as "concrete."

Attachment 8 references the most limiting design for eachsituation.

The analysis has used a bounding treatment for concrete and the methodology remains conservative throughout the life of the concrete.

Additional information is provided inSection B.4.3 of Attachment 8.Biases and Uncertainties Reactivity biases are known variations between the real and analyzed system and theirreactivity impact is added directly to the calculated keff. Uncertainties are random dispersions around a nominal, measured quantity.

Their impact is added to the calculated keff as the squareroot of the sum of the squares of the uncertainties.

The following biases and uncertainties areaccounted for in the analysis.

A detailed discussion of biases and uncertainties is provided inSection B.4.5 of Attachment 8.* Reactivity effect of manufacturing tolerances

• Structural material presence* Eccentric fuel assembly positioning

• Uncertainty in the predictive capability of SCALE 6.1.2 and the associated cross-section library* Temperature bias for operating temperature range* Planar enrichment bias20 Enclosure Description and Assessment of Proposed License Amendment New Fuel Storagqe Rack Criticality Safety AnalysisThe criticality safety analysis for the NFS rack consists of determining the limiting fuel designunder both the fully flooded and optimum moderation condition.

Biases and uncertainties forboth fully flooded and optimum moderation conditions are calculated using the limiting fueldesign. The best estimate keff of the NES rack under both full density water and optimummoderation conditions is less than the target keff. This demonstrates that the NFS rack complieswith the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.6 ofAttachment 8.The analysis of the NES rack has demonstrated that it can be operated in its design capacitywithout risk of exceeding the maximum reactivity imposed by regulation.

The analysis supportsuse of these components up to a maximum radially averaged enrichment of 4.65 weight percentU-235. All fuel used to date at PVNGS has an initial enrichment of < 4.55 weight percent U-235.Intermediate Fuel Storaqe Rack Criticality Safety AnalysisThe criticality safety analysis for the IFSR consists of determining the target keff for the IFSR,then confirming that the best estimate system keff (plus 2 o) is below the target keff with thelimiting fuel design. The analysis uses the NGF design because the NGF is more reactive withfull density water than VAP fuel. The biases and uncertainties are also calculated using the NGFdesign. The best estimate keff of the IFSR is less than the target keff, which demonstrates thatthe IFSR complies with the requirements of 10 CFR 50.68. Additional information is provided inSection B.4.7 of Attachment 8.Fuel Upender.

Transfer

Machine, and New Fuel Elevator Criticality Safety AnalysisThe criticality safety analysis for the fuel upender and transfer

machine, and the analysis for thenew fuel elevator, demonstrate that they can be used with fresh 4.65 weight percent U-235 fuelwithout exceeding a keff of 0.95 at a 95 percent probability, 95 percent confidence level. Thedesign basis fuel is the NGF design. The best estimate keff of the fuel upender and transfermachine, and the new fuel elevator, is less than the target which demonstrates compliance with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.8 ofAttachment 8.3.3 Spent Fuel Pool Transition PlanThe SEP transition will be conducted over a total lapsed time of approximately 24 months with aschedule based on the unit refueling outages.

Therefore, APS will insert two sets of TS and TSBases pages during implementation.

One set will be labeled "Before SEP transition" and theother set will be labeled "After SFP transition."

The Spent Fuel Pool Transition Plan is based on a SEP module-by-module transition scheme.Each SEP module that has not been transitioned is governed by the 'Before SEP transition" pages. As each module in a SEP is transitioned to the new configuration, the Shift Manager willmake an entry in the control room log and declare that module as "transitioned."

That particular module is then governed by the "After SEP transition" pages. When all three units have beentransitioned, APS will submit an administrative TS change to remove the "before" and "after"pages, and insert the final pages.21 Enclosure Description and Assessment of Proposed License Amendment APS will transition to the proposed TS 3.7.17 in each of the three units in the following manner:1. Move fuel assemblies as needed in order to neutronically decouple one module from thebalance of the SEP. Analysis demonstrates that one row of empty cells is enough todecouple modules in the SEP.2. Perform a shuffle of the decoupled module, including installation of NETCO-SNAP-IN rack inserts.a. Some modules may need to be completely emptied of fuel assemblies.

Theassemblies that must be moved may be stored in the appropriate regions of therest of the SEP.b. Fuel assemblies that already meet the new TS 3.17.17 requirements may notneed to be shuffled.

3. Upon completion of the fuel shuffle and NETCO-SNAP-IN rack insert installation, theShift Manager will declare the module has transitioned to the new TS 3.7.17 and enterthis information into the control room log.4. Perform additional fuel shuffles, as needed, in order to move fuel from other parts of theSEP to the recently transitioned module.5. Repeat steps 1 through 4 for all 17 modules.6. Perform a 100 percent pool verification to confirm the following

a. Fuel has been properly moved as confirmed by a 100 percent serial numbercheckb. Stainless steel L-Jnserts are in the locations assumed in the analysis of recordc. NETCO-SNAP-IN rack inserts are in their assumed locations

d. Blocking devices are in their assumed locations A generic Westinghouse study (Reference 6.6) investigated the adequacy of assuming adistance of one cell pitch for neutronic isolation.

The study included NGF, which was determined to be the limiting fuel type used in the PVNGS SEP criticality safety analysis.

The results of thestudy concluded that a distance of approximately 10 cm (3.94 inches) of water was adequate forneutronic decoupling.

Given the required separation distance for neutronic decoupling, fuelassemblies separated by a single cell pitch or more at PVNGS are neutronically decoupled.

Once the Spent Fuel Pool Transition Plan has been started in a particular unit, the insertinstallation and SFP transition shall be executed in a deliberate, safe, and controlled manneruntil complete in that unit. The transition to the new SEP configuration will be completed in allthree units in accordance with the Spent Fuel Pool Transition Plan within two years of the NRCapproval date of the amendment or by December 31, 2019, whichever is later. A regulatory commitment regarding transition plan implementation is provided in Attachment 4.4.0 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements The regulations in 10 CFR 50.36(c)(2)(ii)(B),

Limiting conditions for operation, state:22 Enclosure Description and Assessment of Proposed License Amendment Criterion

2. A process variable, design feature, or operating restriction that is aninitial condition of a design basis accident or transient analysis that eitherassumes the failure of or presents a challenge to the integrity of a fission productbarrier.Technical Specification (TS) 3. 7.17 currently meets this requirement and will continue to meetthis requirement after the proposed changes are approved and implemented.

The regulations in 10 CFR 50.36(c)(4),

Design features, state:Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, wouldhave a significant effect on safety and are not covered in categories described inparagraphs (c)(1), (2), and (3) of this section.TS 4.3.1 currently meets this requirement and will continue to meet this requirement after theproposed changes are approved and implemented.

The regulations in 10 CFR 50.68, Criticality accident requirements, specifically 10 CFR50.68(b)(1) state:Plant procedures shall prohibit the handling and storage at any one time of morefuel assemblies than have been determined to be safely subcritical under themost adverse moderation conditions feasible by unborated water.This requirement is currently met by existing PVNGS fuel handling procedures and will continueto be met by the same procedures after the proposed changes are approved and implemented.

The regulations in 10 CER 50.68(b)(2) state:The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racksare loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or designfeatures prevent such flooding or if fresh fuel storage racks are not used.The regulations in 10 CFR 50.68(b)(3) state:If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racksare assumed to be loaded with fuel of the maximum fuel assembly reactivity and filledwith low-density hydrogenous fluid, the k-effective corresponding to this optimummoderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or designfeatures prevent such moderation or if fresh fuel storage racks are not used.The regulations in 10 CFR 50.68(b)(4) state:If no credit for soluble boron is taken, the k-effective of the spent fuel storageracks loaded with fuel of the maximum fuel assembly reactivity must not exceed0.95, at a 95 percent probability, 95 percent confidence level, if flooded withunborated water. If credit is taken for soluble boron, the k-effective of the spent23 Enclosure Description and Assessment of Proposed License Amendment fuel storage racks loaded with fuel of the maximum fuel assembly reactivity mustnot exceed 0.95, at a 95 percent probability, 95 percent confidence level, ifflooded with borated water, and the k-effective must remain below 1.0(subcritical),

at a 95 percent probability, 95 percent confidence level, if floodedwith unborated water.The requirements in 10 CFR 50.68(b) cited above are met by the nuclear criticality safetyanalyses provided in WCAP- 18030-P.

The results of the criticality analysis form the basis of theproposed TS 3.7.17 changes.

TS 3.7.17 currently meets these requirements and will continue tomeet these requirements after the proposed changes are approved and implemented.

The regulations in 10 CFR 50.68(b)(7) state:The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five(5.0) percent by weight.TS 4.3.1.2 currently meets this requirement and will continue to meet this requirement after theproposed changes are approved and implemented.

The regulations in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear PowerPlants, Criterion 62, Prevention of criticality in fuel storage and handling, state:Criticality in the fuel storage and handling system shall be prevented by physicalsystems or processes, preferably by use of geometrically safe configurations.

TS 3. 7.17 currently meets this requirement and will continue to meet this requirement after theproposed changes are approved and implemented.

The guidance in DSS-ISG-2010-01 is to be used by NRC staff to review nuclear criticality safetyanalyses for the storage of new and spent nuclear fuel as they apply to applications for licenseamendments submitted after September 29, 2011.This license amendment request and WCAP-18030-P were developed using the guidance ofDSS-ISG-2010-01.

4.2 Precedent

The analysis methodology for the site-specific criticality analysis employs the PARAGON code,which is approved for use by the NRC (Reference 6.4).4.:3 Significant Hazards Consideration As required by 10 CFR 50.91(a),

Notice for Public Comment, an analysis of the issue of nosignificant hazards consideration using the standards in 10 CFR 50.92, Issuance ofAmendment, is presented below:1. Does the proposed amendment involve a significant increase in the probability orconsequences of an accident previously evaluated?

Response:

No.24 Enclosure Description and Assessment of Proposed License Amendment The proposed amendment would modify the Palo Verde Nuclear Generating Station(PVNGS) Technical Specifications (TS) to incorporate the results of an updated criticality safety analysis for both new fuel and spent fuel storage.

The revised criticality safetyanalysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previouscriticality safety analysis.

The proposed amendment does not change or modify the fuel, fuel handling processes, number of fuel assemblies that may be stored in the spent fuel pool (SFP), decay heatgeneration rate, or the SFP cooling and cleanup system. The proposed amendment wasevaluated for impact on the following previously evaluated events and accidents:

• fuel handling accident (FHA)* fuel misload event* SEP boron dilution event* seismic event* loss of SEP cooling eventImplementation of the proposed amendment will be accomplished in accordance with theSpent Fuel Pool Transition Plan and does not involve new fuel handling equipment orprocesses.

The radiological source term of the fuel assemblies is not affected by theproposed amendment request.

The EHA radiological dose consequences associated withfuel enrichment at this level are addressed in the PVNGS Updated Final Safety AnalysisReport (UFSAR) Section 15.7.4 and remain unchanged.

Therefore, the proposedamendments do not significantly increase the probability or consequences of a FHA.Operation in accordance with the proposed amendment will not change the probability of afuel misload event because fuel movement will continue to be controlled by approved fuelhandling procedures.

Although there will be additional allowable storage arrays defined bythe amendment, the fuel handling procedures will continue to require identification of theinitial and target locations for each fuel assembly that is moved. The consequences of a fuelmisload event are not changed because the reactivity analysis demonstrates that the samesubcriticality criteria and requirements continue to be met for the limiting fuel misload event.Operation in accordance with the proposed amendment will not change the probability orconsequences of a boron dilution event because the systems and events that could affectSFP soluble boron concentration are unchanged.

The current boron dilution analysisdemonstrates that the limiting boron dilution event will reduce the boron concentration fromthe TS limit of 2150 ppm to 1900 ppm. This leaves sufficient margin to the 1460 ppmcredited by the SFP criticality safety analysis.

The analysis confirms that the time needed fordilution to reduce the soluble boron concentration is greater than the time needed for actionsto be taken to prevent further dilution.

Operation in accordance with the proposed amendment will not change the probability of aseismic event since there are no elements of the updated criticality analysis that influence the occurrence of a seismic event. The consequences of a seismic event are notsignificantly increased because the forcing functions for seismic excitation are not increased and because the mass of storage racks with NETCO-SNAP-IN inserts is not appreciably 25 Enclosure Description and Assessment of Proposed License Amendment increased.

Seismic analyses demonstrate adequate stress levels in the storage racks wheninserts are installed.

Operation in accordance with the proposed amendment will not change the probability of aloss of SEP cooling event because the systems and events that could affect SEP cooling areunchanged.

The consequences are not significantly increased because there are nochanges in the SFP heat load or SEP cooling systems, structures, or components.

Furthermore, conservative analyses indicate that the current design requirements andcriteria continue to be met with the NETCO-S NAP-IN inserts installed.

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

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

Response:

No.The proposed amendment would modify the PVNGS TS to incorporate the results of anupdated criticality safety analysis for both new fuel and spent fuel storage.

The revisedcriticality safety analysis provides an updated methodology that allows credit for neutronabsorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previous criticality safety analysis.

The proposed amendment does not change or modify the fuel, fuel handling processes, number of fuel assemblies that may be stored in the pool, decay heat generation rate, or theSEP cooling and cleanup system. The effects of operating with the proposed amendment are listed below. The proposed amendment was evaluated for the potential of each effect tocreate the possibility of a new or different kind of accident:

• addition of inserts to the SEP storage racks* additional weight from the inserts* new storage patterns* displacement of SEP water by the inserts,Each NETCO-SNAP-IN insert will be placed between a fuel assembly and the storage cellwall, taking up some of the space available on two sides of the fuel assembly.

Analysesdemonstrate that the presence of the inserts does not adversely affect spent fuel cooling,seismic capability, or subcriticality.

The aluminum and boron carbide materials ofconstruction have been shown to be compatible with nuclear fuel, storage racks, and SEPenvironments, and generate no adverse material interactions.

Therefore, placing the insertsinto the SEP storage racks cannot cause a new or different kind of accident.

Operation with the added weight of the NETCO-SNAP-IN inserts will not create a new ordifferent accident.

The analyses of the racks with NETCO-SNAP-IN inserts installed demonstrate that the stress levels in the rack modules continue to be considerably less thanallowable stress limits. Therefore, the added weight from the inserts cannot cause a new ordifferent kind of accident.

26 Enclosure Description and Assessment of Proposed License Amendment Operation with the proposed fuel storage patterns will not create a new or different kind ofaccident because fuel movement will continue to be controlled by approved fuel handlingprocedures.

These procedures continue to require identification of the initial and targetlocations for each fuel assembly that is moved. There are no changes in the criteria ordesign requirements pertaining to fuel storage safety, including subcriticality requirements.

Analyses demonstrate that the proposed storage patterns meet these requirements andcriteria with adequate margins.

Therefore, the proposed storage patterns cannot cause anew or different kind of accident.

Operation with insert movement above stored fuel will not create a new or different kind ofaccident.

The insert with its handling tool weighs less than the weight of a single fuelassembly.

Single fuel assemblies are routinely moved safely over fuel assemblies and thesame level of safety in design and operation will be maintained when moving the inserts.The installed rack inserts will displace a negligible quantity of the SEP water volume andtherefore will not reduce operator response time to previously-evaluated SFP accidents.

The accidents and events previously analyzed remain bounding.

Therefore, the proposedamendment does not create the possibility of a new or different kind of accident from anyaccident previously evaluated.

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

No.The proposed amendment would modify the TS to incorporate the results of an updatedcriticality safety analysis for both new fuel and spent fuel storage.

The revised criticality safety analysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in theprevious criticality safety analysis.

It was evaluated for its effect on current margins of safetyas they relate to criticality, structural integrity, and spent fuel heat removal capability.

The margin of safety for subcriticality required by 10 CFR 50.68(b)(4) is unchanged.

Newcriticality analyses confirm that operation in accordance with the proposed amendment continues to meet the required subcriticality margins.The structural evaluations for the racks and spent fuel pool with NETCO-SNAP-IN insertsinstalled show that the rack and SEP are unimpaired by loading combinations during seismicmotion, and there is no adverse seismic-induced interaction between the rack and NETCO-SNAP-IN inserts.The proposed amendment does not affect spent fuel heat generation, heat removal from thefuel assembly, or the SEP cooling systems.

The effects of the NETCO-SNAP-IN inserts arenegligible with regards to volume of water in the pool, flow in the SEP rack cells, and heatremoval system performance.

The addition of a Spent Fuel Pool Rack Neutron Absorber Monitoring program (proposed TS5.5.21) provides a method to identify potential degradation in the neutron absorber materialprior to challenging the assumptions of the criticality safety analysis related to the material.

Therefore, the addition of this monitoring program does not reduce the margin of safety;27 Enclosure Description and Assessment of Proposed License Amendment rather it ensures th'e margin of safety is maintained for the planned life of the spent fuelstorage racks.Therefore, the proposed amendment does not involve a significant reduction in the marginof safety.4.4 Conclusion APS concludes that operation of the facility in accordance with the proposed amendment doesnot 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 theconsiderations discussed above, (1) there is reasonable assurance that the health and safety ofthe public will not be endangered by operation in the proposed manner, (2) such activities willbe conducted in compliance with the Commission's regulations, and (3) the issuance of theamendment will not be inimical to the common defense and security or the health and safety ofthe public.5.0 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement withrespect to installation or use of a facility component located within the restricted area, as definedin 10 CFR 20, Standards for Protection Against Radiation.

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

offsite, or (iii) asignificant increase in individual or cumulative occupational radiation exposure.

Accordingly, theproposed amendment meets the eligibility criterion for categorical exclusion set forth in10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51 .22(b), no environmental impactstatement or environmental assessment need be prepared in connection with the proposedamendment.

6.0 REFERENCES

6.1 Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3(Proprietary),

WCAP-1 8030-P, Revision 0, September 2015.6.2 Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools,DSS-ISG-201 0-01, Revision 0, Nuclear Regulatory Commission Division of SafetySystems, Rockville, MD, September 29, 2011. (ML1 10620086) 6.3 Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis andDesign, ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge,TN, June 2011.6.4 M. Ouisloumen, H. Huria, et al, Qualification of the Two-Dimensional Transport CodePARAGON, WCAP-16045-P-A, Revision 0, Westinghouse Electric Company LLC,Monroeville, PA, August 2004.28 Enclosure Description and Assessment of Proposed License Amendment 6.5 C. V. Parks, et al, Review and Prioritization of Technical Issues Related to BumupCredit for LWR Fuel, NUREG/CR-6665, Oak Ridge National Laboratory, Oak Ridge,TN, February 2000.6,6 Letter, J. Gresham (WEC) to NRC, Responses to Requests for Additional Information from the Review of WCAP- 1 7483-PA/WCAP-1 7483-NP, Revision 0, 'Westinghouse Methodology for Spent Fuel Pool and New Fuel Rack Criticality Safety Analysis,'

LTR-NRC-15-60, dated July 20, 2015.29 Enclosure

,Description and Assessment of Proposed License Amendment ATTACHMENT 1Marked-up Technical Specifications Pages(Pages Provided for Before and After SEP Transition) 3.7.17 3.7.17-23.7.17-33.7.17-44.0-24.0-35.5-19 SBefore SFP transitionI Spent Fuel Assembly Storage3.7.173.7 PLANT SYSTEMS3.7.17 Spent Fuel Assembly StorageLCO 3.7.17APPLICABILITY:

The combination of initial enrichment, burnup, and decaytime of each fuel assembly stored in each of the fourregions of the fuel storage pool shall be within theacceptable burnup domain for each region as shown in Figures3.7.17-1, 3.7.17-2, or 3.7.17-3, and described inSpecification 4.3.1.1.Whenever any fuel assembly is stored in the fuel storagepool.ACTIONS__________________________

CONDITION REQUIRED ACTION COMPLETION TIMEA. Requirements of the A.1------NOTE----

LCO not met. LCD 3.0.3 is notapplicable.

Initiate action to Immediately move the noncomplying fuel assembly into anappropri ate region.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior toenrichment, burnup, and decay time of the storing thefuel assembly is in accordance with Figures fuel assembly3.7.17-1, 3.7.17-2, or 3.7.17-3, and in the fuelSpecification 4.3.1.1.

storage pool.PALO VERDE UNITS 1,2,33.7.17-IPAL VEDE NIT 1,,3 .7.7-1AMENDMENT NO. 117, 1£ IBefore SFP transition]I Spent Fuel Assembly Storage3.7.17Figure 3.7.17-1ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT forRegion 2060SSSS0ACC EPTABL["

for Reg on 215000500I-NOT A CEPTA LE for R gion 2SSSSSSSSSS0SSSSSSS_________________

_________________

9SSSSSSSSSNote: This curve assumes ero decay time.____________

~1~S(11.52.02.53.0 3.5Initial Enrichment, weight %4.04.5

• 5.04.80%limitingenrichment PALO VERDE UNITS 1,2,33.7.17-2AMENDMENT NO. 117, !2

[Before SFP transition Spent Fuel AssemblyStorage3.7.17Figure 3.7.17-2ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT forRegion 3(at decay times from 0 to 20 years)4500040000350003000025000>20000t'0Eci0ACCEP' ABLE for Region 3 "o,.S//NOT ACq;EPTABI.

E for RegionS0ThOUU+J +

IS0SS(S-a---SS5000~~.2Noe Asnnent and current diS0Scay lime.:ly e/igible for Regi jm 3 if actualIBU

> 3U requirement for given initial endchFi, , , , .....m1.5 2.0 2.5 3.0 3.5 4.0 4.5

• 5.0Initial Enrichment, weight % 4.80%limitingDecaylime[

-U-1-5years

--4--2Oyears enrichment PALO VERDE UNITS 1,2,33..73AMNETNO 3.7.17-3AMENDMENT NO.

I Before SFP transition Spent Fuel Assembly Storage3.7.17Figure 3.7.17-3ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT

~forRegion 4(at decay times from C to 20 years)I-0E~0Ea)0)3.0 3.5Initial Enrichment, weight %limitingDecaylimeI

--0years 11-years

--&.-lOyears

-~-'-15 years --o-20years enrdchment PALO VERDE UNITS 1,2,3 371- MNMN O3.7.17-4AMENDMENT NO.

SAfter SFP transition Spent Fuel Assembly Storage3.7.173.7 PLANT SYSTEMS 13.7.17-1 through 3.7.17-5.I 3.7.17 Spent Fuel Assembly StorageLCO 3.7.11 The combination of initialnichment, burnup, and decaytime of each fuel assemblytreine=hftef'-

acceptable burnup domain for each region as shown in Figures') 7 17 1 ') 7 17 ') ,-~-i Q 7 17 '2 -~A A-,-.~k,-A

-vsSpecification 1.3.1.1.APPLICABILITY:

Whenever any fuelpool.assembly is stored in the fuel storageACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Requirements of the A.1------NOTE--

--LCO not met. LCO 3.0.3 is notapplicable.

Initiate action to Immediately move the noncomplying fuel assembly into anappropriate region.SURVEILLANCE REQUIREMENTS_________

SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior toenrichment, burnup, and decay time of the storing thefuel assembly is in accordance with fuel assembly...and in the fuelSpecification 4mi.3.1.1

... storage pool.ITables 3.7.17-1 through 3.7.17-5, Figure 3.7.17-PALO VERDE UNITS 1,2,33.7.17-1PAL VEDE NIT 1,,3 .7.7-1AMENDMENT NO. I147, 2 Ilnsert new Tables 3.7.17-1 through 3.7.17-5 andjFigure 3.7.17-1 (total 6 pages) here.IIAfter SFP transitionI Spent Fuel Assembly Storage3.7.17FiguJren 3.7.17 1Acc(rkArnl

\J l \lII-1CI IC TkITTT Al FI-dITriUMCIdT PALO VERDE UNITS 1,2,397179PAL VRD UIT 12, 37.72AMENDMENT NO. 11.7, 2 SAfter SFP transition Spent Fuel Assembly Storage3.7.17limiting--me '-O years --U-5 years -* 10 years -U-I--l5 years --4-20 years enrichment PALO VERDE UNITS 1,2,33. 13AMNETNO 1597]79AMENDMENT NO.

SAfter SFP transitionI Spent Fuel Assembly Storage3.7.17Figure,3.7.17 3-ACSE'MILY VU NUP VERSUS TIT-I-AI ENRT/ICHMEITT

'4,~uuuU45000SS40000 __ __ _rSACCEIRegion 4350002500000..~2O0001500010000I 0SS_________________

_________

ISSSS__________________

___________________

S0SSNOT AC E ABLE or Region ____ __SSSSS_________________

_________________

_________________

_________

S______/SSSS0/ ______ + 4-~---f5000//N______ 4- + 4-SS0S000caYthne.

0mq~kemwl fortyB~efor Reg M 4&fachiBU

>-hiaI enddi rdw caidwmt1.2.0 2.5 3.0 3.5 4.0 4.5

• 5Initial Enrichment, weight % 4.80%-ieI-4-0 years --11-5 years 15 years --4-20Oyears lenicmetnt PALO VERDE UNITS 1,2,33..7AMN ETN. §Q717AAMENDMENT NO.

[After SFP transition]

Spent Fuel Assembly Storage3.7.17Table 3.7.17-1Fuel RegionsRanked by Reactivity Fuel Region 1 Highest Reactivity (See Note 2)Fuel Region 2Fuel Region 3Fuel Region 4Fuel Region 5Fuel Region 6 Lowest Reactivity Notes:1. Fuel Regions are defined by assembly average burnup, initial enrichment' and decay time asprovided by Table 3.7.17-2 through Table 3.7.17-5.

2. Fuel Regions are ranked in order of decreasing reactivity, e.g., Fuel Region 2 is less reactivethan Fuel Region 1, etc.3. Fuel Region 1 contains fuel with an initial maximum radially averaged enrichment up to4.65 wt% 235U. No burnup is required.

4. Fuel Region 2 contains fuel with an initial maximum radially averaged enrichment up to4.65 wt% 235U with at least 16.0 GWd/MTU of bumup.5. Fuel Regions 3 through 6 are determined from the minimum burnup (BU) equation andcoefficients provided in Tables 3.7.17-2 through 3.7.17-5.

6. Assembly storage is controlled through the storage arrays defined in Figure 3.7.17-1.

7. Each storage cell in an array can only be populated with assemblies of the Fuel Region definedin the array definition or a lower reactivity Fuel Region.SInitial Enrichment is the nominal 235U enrichment of the central zone region of fuel, excluding axial blankets, priorto reduction in 235U content due to fuel depletion.

If the fuel assembly contains axial regions of different 235Ujenrichment values, such as axial blankets, the maximum initial enrichment value is to be utilized.

[After SFP transition[

Spent Fuel Assembly Storage3.7.17Table 3.7.17-2Fuel Region 3: Burnup Requirement Coefficients Coefficients DecayTime (yr.) A1 A2 A3 A40 -1.5473 15.5395 -39.0197 24.11215 -1.4149 13.9760 -33.6287 18.336910 -1.3012 12.6854 -29.2539 13.687915 -1.0850 10.4694 -22.1380 6.367320 -0.9568 9.1487 -17.9045 2.0337Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meetthe requirements of a Fuel Region, the assembly burnup must exceed the "minimum burnup"(GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated from the following equation:

BU =Al

• En3 + A2

• En2 + A3

• En + A42. Initial enrichment, En, is the maximum radial average 235U enrichment.

Any En value between2.55 wt% 235U and 4.65 wt% 235U may be used. Burnup credit is not required for an En below2.55 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decaytime of greater than 20 years.

[After SFP transition Spent Fuel Assembly Storage3.7.17Table 3.7.17-3Fuel Region 4: Burnup Requirement Coefficients Coefficients DecayTime (yr.) A1 A2 A3 A40 0.4260 -6.2766 40.9264 -54.68135 0.2333 -4.1!545 32.9080 -46.116110 0.4257 -6.2064 39.0371 -51.588915 0.53 15 -7.3777 42.5706 -54.752420 0.5222 -7.3897 42.6587 -54.8201Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meetthe requirements of a Fuel Region, the assembly bumup must exceed the "minimum burnup"(GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated from the following equation:

BU=AI

• En3 +A2 *En2+-bA3 *En +A42. Initial enrichment, En, is the maximum radial average 235U enrichment.

Any En value between1.75 wt% 235U and 4.65 wt% 235U may be used. Bumnup credit is not required for an En below1.75 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decaytime of greater than 20 years.

[After SEP transition Spent Fuel Assembly Storage3.7.17Table 3.7.17-4Fuel Region 5: Burnup Requirement Coefficients Decay Coefficients Time(yr.) A1 A2 A3 A40 -0.1114 -0.4230 20.9136 -32.85515 -0.1232 -0.4463 20.8337 -32.606810 -0.2357 0.4892 18.0192 -30.004215 -0.1402 -0.4523 20.3745 -31.756520 -0.0999 -0.8152 21.0059 -31.9911Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly tomeet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum bumup"(GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated from the following equation:

BU =A1

• En3 + A2

• En2 + A3

• En + A42. Initial enrichment, En, is the maximum radial average 235U enrichment.

Any En value between1.65 wt% 235U and 4.65 wt% 235U may be used. Burnup credit is not required for an En below1.65 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decaytime of greater than 20 years.

[After SFP transition Spent Fuel Assembly Storage3.7.17Table 3.7.17-5Fuel Region 6: Burnup Requirement Coefficients Decay Coefficients Time(yr.) A1 A2 A3 A40 0.7732 -9.3583 49.6577 -54.68475 0.7117 -8.4920 45.1124 -49.728210 0.6002 -7.2638 40.2603 -44.934815 0.5027 -6.2842 36.6715 -41.493420 0.2483 -3.7639 28.8269 -34.6419Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

For instance, no additional allowance for bumnup uncertainty or enrichment uncertainty is required.

For a fuel assembly tomeet the requirements of a Fuel Region, the assembly bumup must exceed the "minimum burnup"(GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated from the following equation:

BU =A1

• En3 + A2

• En2 + A3

• En + A42. Initial enrichment, En, is the maximum radial average 235U enrichment.

Any En value between1.45 wt% Z3U and 4.65 wt% 23U may be used. Burnup credit is not required for an En below1.45 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decaytime of greater than 20 years.

SAfter SFP trransition Spent Fuel Assembly Storage3.7.17Figure 3.7.17-1Allowable Storage ArraysFwou Region 6 assemblies (6) Tw steckrbagded cells cotain abtlocess stells L(ise). TheRgo1. Thsembishaded loathionsa indicatehelhcoti a stainless steel L-insert.NoETOSA-N 2.o Ae block1asedbis()cekrbaddwt w cells(X contains lcing dvcanolyw terainshe actiefulrein 3TC. NTheRgo-n1assemlNinets must bc oientedl winthe saedrcinaah stainless steel L-inserts.Eer 4 NTC -N PN isrsaeolloaeincells without a stainless steel L-inetms oti EC -N P1insert.

5-nsr. Anyhel egontann3 afe assemblyyr iCsa iseaiennemt (aerflld cell intinn aNTCall-Niset stoageayrys 6.e AnR traearagoaion deintdfrafe assembly may cbeckreplaced with noren-gon4fssssiele).Th mein2atserial.

n h ignlylctdRein4asml r ahi I Before SFP transitionI Design Features4.04.0 DESIGN FEATURES (continued) 4.3 Fuel Storage4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall bemaintained with:a. Fuel assemblies having a maximum radially averagedU-235 enrichment of 4.80 weight percent;b. keff < 1.0 if fully flooded with unborated water,which includes an allowance for biases anduncertainties as described in Section 9.1 of theUFSAR;c. keff 0.95 if fully flooded with water borated to900 ppm, which includes an allowance for biases anduncertainties as described in Section 9.1 of theUFSAR.d. A nominal 9.5 inch center-to-center distancebetween adjacent storage cell locations.

e. Region 1: Fuel shall be stored in a checkerboard (two-out-of-four) storage pattern.

Fuel thatqualifies to be stored in Regions 1, 2, 3, or 4 inaccordance with Figures 3.7.17-1, 3.7.17-2, or3.7.17-3, may be stored in Region 1.f. Region 2: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2(two-out-of-twelve) assemblies and Region 4(ten-out-of-twelve) assemblies are mixed as shownin Section 9.1 of the UFSAR. Only fuel thatqualifies to be stored in Regions 2, 3, or 4, inaccordance with Figures 3.7.17-1, 3.1.17-2, or3.7.17-3, may be stored in Region 2.g. Region 3: Fuel shall be stored in a four-out-of-four storage pattern.

Only fuel that qualifies tobe stored in Regions 3 or 4, in accordance withFigures 3.7.17-2 or 3.7.17-3, may be stored inRegion 3.(conti nued)PALO VERDE UNITS 1,2,34.0-2PALOVERE UITS1.23 40-2AMENDMENT NO. 47~

I Before SEP transition IDesign Features4.04.0 DESIGN FEATURES (continued)

h. Region 4: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2(two-out-of-twelve) assemblies and Region 4(ten-out-of-twelve) assemblies are mixed as shownin Section 9.1 of the UFSAR. Only fuel thatqualifies to be stored in Region 4 in accordance with Figure 3.7.17-3 shall be stored in Region 4.4.3.1.2 The new fuel storage racks are designed and shall bemaintained with:a. Fuel assemblies having a maximum radially averagedU-235 enrichment of 4.80 weight percent;b. keff 0.95 if fully flooded with unborated water,which includes an allowance for biases anduncertainties as described in Section 9.1 of theUFSAR:c. keff 0.98 if moderated by aqueous foam, whichincludes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR; andd. A nominal 17 inch center to center distance betweenfuel assemblies placed in the storage racks.4.3.2 DrainageThe spent fuel storage pool is designed and shall be maintained toprevent inadvertent draining of the pool below elevation 137 feet -6 inches.4.3.3 CapacityThe spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuelassemblies.

PALO VERDE UNITS 1,2,34.0-3PAO EDEUNT 12, .03AMENDMENT NO. 11 o ...

After SFP transition Design Features4.04.0 DESIGN FEATURES (continued) 4.3 Fuel Storage4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall bemaintained with:46a. Fuel assemblies having/ maximum radially averagedU-235 enrichment .weight percent:b. keff < 1.0 if fully flooded with unborated water.which includes an allowance for biases anduncertainties as described in Section 9.1 of theUFSAR-c. keff -< 0.95 if fully flooded with water borated to~ppm, which includes an allowance for biases andunertainties as described in Section 9.1 of the1460 UFSAR.d. A nominal 9.5 inch center-to-center distancebetween adjacent storage cell locations.

e. R~on !:Fuel shall bc stored in a checkerboa=rd Fuel assemblies are .. ..... --.classified in Fuel Regions qualifie to be tore in Regon 1 2 3, or in1-6 as shown in Tables a ccordance w.i th Figu=res 3.7.17 1, 3.7.17 2, or3.7.17-1 through 3. 7.17 3 .. mayb in Rcgn 1.3.7.17-5. P Fe,,l shall be a; .......tin-3 byIstrg pattern in ..hich Regon 2( .. = atenrutontele seble arfel mixed as. shownaccordance wit FiguQres 3.7.17 1, 3.7/.17%

2,-n orRegionnu3.

PALO VERDE UNITS 1,2,340-AMNETNO

.&4.0-2AMENDMENT NO. 117, 125 After SFP transition Design Featuresi 4.04.0 DESIGN FEATURES (continued) 4.3.1.2 Th enewfuel Fuorag rhalls bre dstored ind sa llcat be3ai bta ine witorgh ateni-hch go(t.oF uteflwlc assemblies hvn m a ndmu ra gionl avrae(t35enrouioctele)t afswembieht aerenmxdt shwincerectiones9.1 dof cthe edUF n S. tion 9.1& of thqualifies to blowne store binaRegio and uccordaincies ait Fiuesrib3 hllb soed in Regtion 9.Ifte FA nm.Ainta ined with inhc, r 465t ene itnc ewea.fuel assemblies hlavin maximumoradill raverageU-235raiagientrichmenteor eih eretThesp bn kefue strg 0.95 if fullyflooed withalunboaed waitaiert preveninawhichn drincldeing allowanepfor biaswesevatind u3 et nchertitessdsrbd nScin91o hThespent includstoane allowanceforsbiase and uncl ermitaintes withstoasgescrpaibyliied iSetiono 9.1ofe then UF3AR andlfuesaseblesplce hisorgeraks PALO VERDE UNITS 1,2,34.0-3PAL VRD UITS1,.34.-3AMENDMENT NO. 1-17,12 ..

IBefore SFP transitionI Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)

4. In Regulatory Guide 1.129, Regulatory Position 3,Subsection 5.4.1, "State of Charge Indicator,"

thefollowing statements in paragraph (d) may be omitted:"When it has been recorded that the charging current hasstabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially highcharging current decreases sharply and the battery voltagerises to approach the charger output voltage."

5. In lieu of RG 1.129. Regulatory Position 7, Subsection 7.6, "Restoration,"

the following may be used: "Following the test, record the float voltage of each cell of thestring."b. The program shall include the following provisions:

1. Actions to restore battery cells with float voltage<2.13 V;2. Actions to determine whether the float voltage of theremaining battery cells is 2.13 V when the floatvoltage of a battery cell has been found to be<2.13 V:3. Actions to equalize and test battery cells that hadbeen discovered with electrolyte level below the topof the plates:4. Limits on average electrolyte temperature, batteryconnection resistance, and battery terminal voltage;and5. A requirement to obtain specific gravity readings ofall cells at each discharge test, consistent withmanufacturer recommendations.

PALO VERDE UNITS 1,2,3 551 MNMN O 95.5-19AMENDMENT NO.

IAfter SFP TransitionI Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)

4. In Regulatory Guide 1.129. Regulatory Position 3,Subsection 5.4.1, "State of Charge Indicator."

thefollowing statements in paragraph (d) may be omitted:"When it has been recorded that the charging current hasstabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially highcharging current decreases sharply and the battery voltagerises to approach the charger output voltage."

5. In lieu of RG 1.129, Regulatory Position 7, Subsection 1.6. "Restoration."

the following may be used: "Following the test, record the float voltage of each cell of thestring."b. The program shall include the following provisions:

1. Actions to restore battery cells with float voltage<2.13 V:2. Actions to determine whether the float voltage of theremaining battery cells is 2.13 V when the floatvoltage of a battery cell has been found to be<2.13 V:3. Actions to equalize and test battery cells that hadbeen discovered with electrolyte level below the topof the plates:4. Limits on average electrolyte temperature, batteryconnection resistance, and battery terminal voltage:and5. A requirement to obtain specific gravity readings ofall cells at each discharge test, consistent withmanufacturer recommendations.

page 5.5-19PALO VERDE UNITS 1,2.3 551 MNMN O 95.5-19AMENDMENT NO.

[After SFP Transition]

Insert for page 5.5-195.5.21 Spent Fuel Storagqe Rack Neutron Absorber Monitoring ProgqramCertain storage cells in the spent fuel storage racks utilize neutron absorbing materialthat is credited in the spent fuel storage rack criticality safety analysis to ensure thelimitations of Technical Specifications 3.7.17 and 4.3.1.1 are maintained.

In order to ensure the reliability of the neutron absorber

material, a monitoring programis provided to confirm the assumptions in the spent fuel pool criticality safety analysis.

The Spent Fuel Storage Rack Neutron Absorber Monitoring Program shall requireperiodic inspection and monitoring of spent fuel pool test coupons and neutron absorberinserts on a performance-based frequency, not to exceed 10 years.Test coupons shall be inspected as part of the monitoring program.

These inspections shall include visual, B-10 areal density and corrosion rate.Visual in-situ inspections of inserts shall also be part of the program to monitor for signsof degradation.

In addition, an insert shall be removed periodically for visual inspection, thickness measurements, and determination of retention force.

Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 2Revised Technical Specifications Pages (Clean Copy)(Pages Provided for Before and After SEP Transition) 3.7.17-13.7.17-23.7.17-33.7.17-43.7.17-53.7.17-63.7.17-74.0-24.0-35.5-195.5-20 Before SFP transition Spent Fuel Assembly Storage3.7.173.7 PLANT SYSTEMS3.7.17 Spent Fuel Assembly StorageLCO 3.7.17The combination of initial enrichment, burnup, and decaytime of each fuel assembly stored in each of the fourregions of the fuel storage pool shall be within theacceptable burnup domain for each region as shown in Figures3.7.17-1, 3.7.17-2, or 3.7.17-3, and described inSpecification 4.3.1.1.APPLICABILITY:

Whenever any fuelpool.assembly is stored in the fuel storageACTIONS ________________

CONDITION REQUIRED ACTION COMPLETION TIMEA. Requirements of the A.1------NOTE----

LCO not met. LCO 3.0.3 is notapplicable.

Initiate action to Immediately move the noncomplying fuel assembly into anappropriate region.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior toenrichment, burnup, and decay time of the storing thefuel assembly is in accordance with Figures fuel assembly3.7.17-1, 3.7.17-2, or 3.7.17-3, and in the fuelSpecification 4.3.1.1.

storage pool.PALO VERDE UNITS 1,2,3 371- MNMN O -~3.7.17-1AMENDMENT NO.

Before SFP transition Spent Fuel Assembly Storage3.7.17Figure 3.7.17-1ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT forRegi on 220000EE1.5 2.0 2.5 3.03.54.0Initial Enrichment, weight %4.5

• 5.04.80%limitingenrichment PALO VERDE UNITS 1,2,3 371- MNMN O3.7.17-2AMENDMENT NO.

Before SFP transition Spent Fuel Assembly Storage3.7.17Figure 3.7.17-2ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT forRegi on 3(at decay times from 0 to 20 years)Eci1.5 2.0 2.5 3.0 3.5 4.0 4.5

• 5.0Initial Enrichment, weight % 4.80%--q Oyers --m-5yars l~ears --X-15ear -- --2yeas ....limitingflmj --Oyars -U-yers ~-1yeas -E-l~yars oyers enrichment DecaylPALO VERDE UNITS 1,2,3 371- MNMN O 23.7.17- 3AMENDMENT NO.

Before SFP transition Spent Fuel Assembly Storage3.7.17Figure 3.7.17-3ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT forRegi on 4(at decay times from 0 to 20 years)5000045000BSA14 + -~ -S'IvuuU ______________

ACCE TABLE fo Region 435000 ____________

30000BBS~.nnn ___________

I-E 20000(I)(0J,I)150001000050000-1.5JSSSSSNOT ACC EPTABLE ~or Region j4/ -U--BSii~-i I SBB___ I ___ ___ ___ ___ __ B-#A~-- 4 + + +/.JSSSBS---S0BSBSSBBcaytime.:

+ t A------N+ote: Assern4Iy eflgbJ for Regi in 4 facua IBU>* U requiefent forjiI ghen irai ech ent and currnt d2-.0 2.5 3.0 3.5 4.0 4.5 a 5.0Initial Enrichment, weight % 4.80%limitinga -U-!--5 years -~-k--l years --UP-15 years --O-20 years .Jenrichment Decaylimej ye.PALO VERDE UNITS 1,2,3 371- MNMN O 23.7.17-4AMENDMENT NO.

After SFP transition ISpent Fuel Assembly Storage3.7.173.7 PLANT SYSTEMS3.7.17 Spent Fuel Assembly StorageLCO 3.7.17The combination of i niti al enrichment, burnup, and decaytime of each fuel assembly shall be in compliance with therequirements specified in Tables 3.7.17-1 through 3.7.17-5.

APPLICABILITY:

Whenever any fuelpool.assembly is stored in the fuel storageACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Requirements of the A.1------NOTE----

LCO not met. LCO 3.0.3 is notapplicable.

Initiate action to Immediately move the noncomplying fuel assembly into anappropriate region.SURVEILLANCEREQUIREMENTS__________

SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior toenrichment, burnup, and decay time of the storing thefuel assembly is in accordance with Tables fuel assembly3.7.17-1 through 3.7.17-5, Figure 3.7.17-1, in the fueland Specification 4.3.1.1.

storage pool.PALO VERDE UNITS 1,2,337171AEDNT O.I,3.7.17-1AMENDMENT NO.

After SFP transition Spent Fuel Assembly Storage3.7.17Table 3.7.17-1Fuel RegionsRanked by Reactivity Fuel Region 1 Highest Reactivity (See Note 2)Fuel Region 2Fuel Region 3Fuel Region 4Fuel Region 5Fuel Region 6 Lowest Reactivity Notes:1. Fuel Regions are defined by assembly average burnup. initial enrichment' and decaytime as provided by Table 3.7.11-2 through Table 3.7.17-5.

2. Fuel Regions are ranked in order of decreasing reactivity, e.g.. Fuel Region 2 isless reactive than Fuel Region 1. etc.3. Fuel Region 1 contains fuel with an initial maximum radially averaged enrichment upto 4.65 wt% 235U. No burnup is required.

4. Fuel Region 2 contains fuel with an initial maximum radially averaged enrichment upto 4.65 wt% 235U with at least 16.0 GWd/MTU of burnup.5. Fuel Regions 3 through 6 are determined from the minimum burnup (BU) equation andcoefficients provided in Tables 3.7.17-2 through 3.7.17-5.

6. Assembly storage is controlled through the storage arrays defined in Figure 3.7.17-1.

7. Each storage cell in an array can only be populated with assemblies of the FuelRegion defined in the array definition or a lower reactivity Fuel Region.'Initial Enrichment is the nominal 235U enrichment of the central zone region of fuel, excluding axialblankets, prior to reduction in 2350 content due to fuel depletion.

If the fuel assembly contains axialregions of different 235U enrichment values, such as axial blankets, the maximum initial enrichment value is to be utilized.

PALO VERDE UNITS 1,2,337.72AEDNTO.1, 3.7.17-2AMENDMENT NO.

After SEP transition ISpent Fuel Assembly Storage3.7.17Table 3.7.17-2Fue] Region 3: Burnup Requi rement Coefficients Decay Coefficients Time (yr.) Ai NmA A40 -1.5473 15.5395 -39.0197 24.11215 -1.4149 13.9760 -33.6287 18.336910 -1.3012 12.6854 -29.2539 13.687915 -1.0850 10.4694 -22.1380 6.367320 -0.9568 9.1487 -17.9045 2.0337Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

Forinstance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meet the requirements of a FuelRegion, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given bythe curve fit for the assembly "decay 'time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated fromthe following equation:

BU = Ai

• En3 + A2

• En2 + A3

• En + A42. Initial enrichment, En, is the maximum radial average 23enrichment.

Any Envalue between 2.55 wt% 235U and 4.65 wt% 235U may be used. Burnup credit is notrequired for an En below 2.55 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based ondecay time.4. The 20-year coefficients must be used to calculate the minimum BU for anassembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN O 23.7.17-3AMENDMENT NO.

After SFP transition Spent Fuel Assembly Storage3.7.17Table 3.7.17 -3Fue] Region 4: Burnup Requirement Coefficients Decay Coefficients Time (yr.) Ai #0 0.4260 -6.2766 40.9264 -54.68135 0.2333 -4.1545 32.9080 -46.116110 0.4257 -6.2064 39.0371 -51.588915 0.5315 -7.3777 42.5706 -54.752420 0.5222 -7.3897 42.6587 -54.8201Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

Forinstance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meet the requirements of a FuelRegion, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given bythe curve fit for the assembly "decay time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated fromthe following equation:

BU =Ai

• En3 + A2

• En2 + A3

• En + A42. Initial enrichment, En, is the maximum radial average 235U enrichment.

Any Envalue between 1.15 wt% 235U and 4.65 wt% 23may be used. Burnup credit is notrequired for an En below 1.75 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based ondecay time.4. The 20-year coefficients must be used to calculate the minimum BU for anassembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN O 23.7.17-4AMENDMENT NO.

After SFP transition ISpent Fuel Assembly Storage3.7.17Table 3.7.17 -4Fue] Region 5: Burnup Requirement Coefficients Decay Coeffi ci entsTime (yr.) Ai km# A40 -0.1114 -0.4230 20.9136 -32.85515 -0. 1232 -0. 4463 20. 8337 -32. 606810 -0. 2357 0.4892 18. 0192 -30. 004215 -0.1402 -0. 4523 20. 3745 -31. 756520 -0. 0999 -0. 8152 21. 0059 -31. 9911Notes :1. Relevant uncertainties are explicitly included in the criticality analysis.

Forinstance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meet the requirements of a FuelRegion, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given bythe curve fit for the assembly "decay time" and "initial enrichment."

Thespeci fi c minimum burnup (BU) requi red for each fuel assembly is calculated fromthe fol lowing equati on:BU= Ai

• En3 + A2

• En2 + A3

• En + A42. Initial enrichment, En. is the maximum radial average 235U enrichment.

Any Envalue between 1.65 wt% 235U and 4.65 wt% 235U may be used. Burnup credit is notrequired for an En below 1.65 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based ondecay time.4. The 20-year coefficients must be used to calculate the minimum BU for anassembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN D3.7.17-5AMENDMENT NO.

After SFP transition Spent Fuel Assembly Storage3.7.17Table 3.7.17 -5Fuel Region 6: Burnup Requirement Coefficients Decay Coefficients Time (yr.) A A2 A4 A0 0.7732 -9.3583 49.6577 -54.68475 0.7117 -8.4920 45.1124 -49.728210 0.6002 -7.2638 40.2603 -44.934815 0.5027 -6.2842 36.6715 -41.493420 0.2483 -3.7639 28.8269 -34.6419Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

Forinstance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meet the requirements of a FuelRegion, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given bythe curve fit for the assembly "decay time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated fromthe fol lowi ng equati on:BU =A1

• En3 + Am

• En2 + A3

• En + A42. Initial enrichment, En. is the maximum radial average 235U enrichment.

Any Envalue between 1.45 wt% 235U and 4.65 wt% 235U may be used. Burnup credit is notrequired for an En below 1.45 wt% 2350.3. It is acceptable to linearly interpolate between calculated BU limits based ondecay time.4. The 20-year coefficients must be used to calculate the minimum BU for anassembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN O3.7.17-6AMENDMENT NO.

After SFP transition Spent Fuel Assembly Storage3.7.17Figure 3.7.17-1Allowable Storage ArraysArray A 1 XTwo Region 1 assemblies (1) checkerboarded with two blocked cells (X).The Region 1 assemblies are each in a cell with a stainless steelL-insert.

No NETCO-SNAP-IN inserts are credited.X 1Array B 1 TCTwo Region 1 assemblies (1) checkerboarded with two cells containing trash cans (TC). The Region 1 assemblies are each in a cell with astainless steel L-insert.

Every cell without a stainless steel L- TC Iinsert must contain a NETCO-SNAP-IN insert.Array CTwo Region 2 assemblies (2) checkerboarded with one Region 3 assembly 2 X(3) and one blocked cell (X). The Region 2 assemblies are each in a ....... ..............

cell with a stainless steel L-insert.

The Region 3 assembly is in a 3 2cell containing a NETCO-SNAP-IN insert.Array 0One Region 2 assembly (2) checkerboarded with three Region 4 2 4assemblies (4). The Region 2 assembly and the diagonally located ..Region 4 assembly are each in a storage cell with a stainless steel L-insert. The two storage cells without a stainless steel L-insert 4 4contain a NETCO-SNAP-IN insert.Array E 5 5Four Region 5 assemblies (5). Two storage cells contain a stainless steel L-insert.

One cell contains a NETCO-SNAP-IN insert. One5storage cell contains no insert.5 5Array F 6 6Four Region 6 assemblies (6). Two storage cells contain a stainless steel L-insert.

The other two cells contain no inserts.

6 6Notes-1. The shaded locations indicate cells which contain a stainless steel [-insert.

2. A blocked cell (X) contains a blocking device and only water in the active fuel region.3. NETCO-SNAP-IN inserts must be oriented in the same direction as the stainless steel [-inserts.

4. NETCO-SNAP-IN inserts are only located in cells without a stainless steel [-insert.

5. Any cell containing a fuel assembly or a TC may instead be an empty (water-filled) cell inall storage arrays.6.Any storage array location designated for a fuel assembly may be replaced with non-fissile material.

PALO VERDE UNITS 1,2,3 371- MNMN O3.7.17-7AMENDMENT NO.

Before SFP transition Design Features4.04.0 DESIGN FEATURES (continued) 4.3 Fuel Storage4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall bemaintained with:a. Fuel assemblies having a maximum radially averagedU-235 enrichment of 4.80 weight percent;b. keff < 1.0 if fully flooded with unborated water,which includes an allowance for biases anduncertainties as descri bed i n Secti on 9.1I of theUFSAR;c. keff -< 0.-95 if fully flooded with water borated to900 ppm, which includes an allowance for biases anduncertainties as described in Section 9.1 of theUFSAR.d. A nominal 9.5 inch center-to-center distancebetween adjacent storage cell locations.

e. Region 1: Fuel shall be stored in a checkerboard (two-out-of-four) storage pattern.

Fuel thatqualifies to be stored in Regions 1, 2, 3, or 4 inaccordance with Figures 3.7.17-1, 3.7.17-2, or3.7.17 -3, may be stored in Region 1.f. Region 2: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2(two-out-of-twelve) assemblies and Region 4(ten-out-of-twelve) assemblies are mixed as shownin Section 9.1 of the UFSAR. Only fuel thatqualifies to be stored in Regions 2, 3, or 4, inaccordance with Figures 3.7.17-1, 3.7.17-2, or3.7.17-3, may be stored in Region 2.g. Region 3: Fuel shall be stored in a four-out-of-four storage pattern.

Only fuel that qualifies tobe stored in Regions 3 or 4, in accordance withFigures 3.7.17-2 or 3.7.17-3, may be stored inRegi on 3.(conti nued)PALO VERDE UNITS 1,2,3 402AEDETN.1k 4.0-2AMENDMENT NO.

Before SEP transition Design Features4.04.0 DESIGN FEATURES (continued)

h. Region 4: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2(two-out-of-twelve) assemblies and Region 4(ten-out-of-twelve) assemblies are mixed as shownin Section 9.1 of the UFSAR. Only fuel thatqualifies to be stored in Region 4 in accordance with Figure 3.7.17-3 shall be stored in Region 4.4.3.1.2 The new fuel storage racks are designed and shall bemaintained with:a. Fuel assemblies having a maximum radially averagedU-235 enrichment of 4.80 weight percent;b. keff 0.95 if fully flooded with unborated water,which includes an allowance for biases anduncertainties as described in Section 9.1 of theUFSAR:c. keff 0.98 if moderated by aqueous foam, whichincludes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR; andd. A nominal 17 inch center to center distance betweenfuel assemblies placed in the storage racks.,.4.3.2 DrainageThe spent fuel storage pool is designed and shall be maintained toprevent inadvertent draining of the pool below elevation 137 feet -6 inches.4.3.3 CapacityThe spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuelassemblies.

PALO VERDE UNITS 1,2,3 403AEDETN.I~

4.0-3AMENDMENT NO.

After SFP transition Design Features4.04.0 k[DESIGN FEATURES(conti nued)4.3 Fuel Storage4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall bemai ntai ned with :a. Fuel assemblies having a maximum radially averagedU-235 enrichment of 4.65 weight percent;b. keff < 1.0 if fully flooded with unborated water,which includes an allowance for biases anduncertai nti es as descri bed in Secti on 9.1 of theUFSAR;c. keff -< 0.95 if fully flooded with water borated to1460 ppm, which includes an allowance for biasesand uncertainties as described in Section 9.1 ofthe UFSAR.d. A nominal 9.5 inch center-to-center distancebetween adjacent storage cell locations.

e. Fuel assemblies, are classified in Fuel Regions 1-6as shown in Tables 3.7.17-1 through 3.7.17-5.

(conti nued)PALO VERDE UNITS 1,2,340-AMNETNO 4.0-2AMENDMENT NO.

After SFP transition Design Features4.04.0 DESIGN FEATURES (continued) 4.3.1.2 The new fuel storage racks are designed and shall bemaintained with:a. Fuel assemblies having a maximum radially averagedU-235 enrichment of 4.65 weight percent;b. keff 0.95 if fully flooded with unborated water,which includes an allowance for biases anduncertai nti es as descri bed i n Secti on 9.1I of theU FSAR;c. keff 0.98 if moderated by aqueous foam, whichi ncl udes an all owance for bi ases and uncertai nti esas described in Section 9.1 of the UFSAR; andd. A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center distance between fuelassemblies placed in the storage racks.4.3.2 DrainageThe spent fuel storage pool is designed and shall be maintained toprevent inadvertent draining of the pool below elevation 137 feet -6 inches.4.3.3 CapacityThe spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuelassemblies.

PALO VERDE UNITS 1,2,340-AMNETNO 2g4.0-3AMENDMENT NO.

Before SFP transition Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)

4. In Regulatory Guide 1.129, Regulatory Position 3,Subsection 5.4.1, "State of Charge Indicator,"

thefollowing statements in paragraph (d) may be omitted:"When it has been recorded that the charging current hasstabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially highcharging current decreases sharply and the battery voltagerises to approach the charger output voltage."

5. In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration,"

the following may be used: "Following the test, record the float voltage of each cell of thestring."b. The program shall include the following provisions:

1. Actions to restore battery cells with float voltage<2.13 V;2. Actions to determine whether the float voltage of theremaining battery cells is 2.13 V when the floatvoltage of a battery cell has been found to be<2.13 V;3. Actions to equalize and test battery cells that hadbeen discovered with electrolyte level below the topof the plates;4. Limits on average electrolyte temperature, batteryconnection resistance, and battery terminal voltage;and5. A requirement to obtain specific gravity readings ofall cells at each discharge test, consistent withmanufacturer recommendations.

PALO VERDE UNITS 1,2,3 551 MNMN O 95.5-19AMENDMENT NO.

After SFP transition Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)

4. In Regulatory Guide 1.129, Regulatory Position 3,Subsection 5.4.1, "State of Charge Indicator,"

thefollowing statements in paragraph (d) may be omitted:"When it has been recorded that the charging current hasstabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially highcharging current decreases sharply and the battery voltagerises to approach the charger output voltage."

5. In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration,"

the following may be used: "Following the test, record the float voltage of each cell of thestring."b. The program shall include the following provisions:

1. Actions to restore battery cells with float voltage<2.13 V;2. Actions to determine whether the float voltage of theremaining battery cells is 2.13 V when the floatvoltage of a battery cell has been found to be<2.13 V;3. Actions to equalize and test battery cells that hadbeen discovered with electrolyte level below the topof the plates;4. Limits on average electrolyte temperature, batteryconnection resistance, and battery terminal voltage;and5. A requirement to obtain specific gravity readings ofall cells at each discharge test, consistent withmanufacturer recommendations.

PALO VERDE UNITS 1,2,35519AEDNT O.495.5-19AMENDMENT NO.

After SFP transition Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.21 Spent Fuel Storage Rack Neutron Absorber Monitoring ProgramCertain storage cells in the spent fuel storage racks utilizeneutron absorbing material that is credited in the spent fuelstorage rack criticality safety analysis to ensure thelimitations of Technical Specifications 3.7.17 and 4.3.1.1 aremaintained.

In order to ensure the reliability of the neutron absorbermaterial, a monitoring program is provided to confirm theassumptions in the spent fuel pool criticality safety analysis.

The Spent Fuel Storage Rack Neutron Absorber Monitoring Programshall require periodic inspection and monitoring of spent fuelpool test coupons and neutron absorber inserts on aperformance-based frequency, not to exceed 10 years.Test coupons shall be inspected as part of the monitoring program.

These inspections shall include visual, B-IO arealdensity and corrosion rate.Visual in-situ inspections of inserts shall also be part of theprogram to monitor for signs of degradation.

In addition, aninsert shall be removed periodically for visual inspection, thickness measurements, and determination of retention force.PALO VERDE UNITS 1,2,35520AEDNT O.195.5-20AMENDMENT NO.

Enclosure Description and Assessment of Proposed License Amendment ATTACHM ENT 3Marked-up Technical Specifications Bases Pages(Pages Provided for Before and After SEP Transition)

B 3.7.15-1B 3.7.15-2B 3.7.17-1B 3.7.17-2B 3.7.17-3B 3.7.17-4B 3.7.17-5B 3.7.17-6C IBefore SFP transitionJ Fuel Storage Pool Boron Concentration B 3.7.15B 3.7 PLANT SYSTEMSB 3.7.15 Fuel Storage Pool Boron Concentration BASESBACKGROUND As described in LCO 3.7.17, 'Spent Fuel Assembly Storage,"

fuel assemblies are stored in the spent fuel racks inaccordance with criteria based on initial enrichment anddischarge burnup. Although the water in the spent fuel poolis normally borated to >_ 2150 ppm, the criteria that limitthe storage of a fuel assembly to specific rack locations isconservatively developed without taking credit for boron.In order to maintain the spent fuel pool keff < 1.0, asoluble boron concentration of 900 ppm is required tomaintain the spent fuel pool keff _< 0.95 assuming the mostlimiting single fuel mishandling accident.

APPLICABLE SAFETY ANALYSESA fuel assembly could be inadvertently loaded into a spentfuel rack location not allowed by LCO 3.7.17 (e.g., anunirradiated fuel assembly or an insufficiently depletedfuel assembly).

Another type of postulated accident isassociated with a fuel assembly that is dropped onto thefully loaded fuel pool storage rack or between a rack andthe pool walls. These incidents could have a positivereactivity effect, decreasing the margin to criticality.

However, the negative reactivity effect of the soluble boroncompensates for the increased reactivity caused by thesepostulated accident scenarios.

The concentration of dissolved boron in the fuel poolsatisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCOThe specified concentration of dissolved boron in the fuelpool preserves the assumptions used in the analyses of thepotential accident scenarios described above. Thisconcentration of dissolved boron is the minimum requiredconcentration for fuel assembly storage and movement withinthe fuel pool.APPLICABILITY This LCO applies whenever any fuel assembly is stored inthe spent fuel pool in order to comply with theTS 4.3.1.1.c design requirement that keff 0.95.(conti nued)PALO VERDE UNITS 1,2,3B37.51RVSO B 3.7.15-1REVISION 3

Before SFP transition Fuel Storage Pool Boron Concentration B 3.7.15BASES (continued)

ACTIONSA.1 and A.2The Required Actions are modified by a Note indicating thatLCO 3.0.3 does not apply.When the concentration of boron in the spent fuel pooi isless than required, immediate action must be taken topreclude an accident from happening or to mitigate theconsequences of an accident in progress.

This is mostefficiently achieved by immediately suspending the movementof fuel assemblies.

This does not preclude the movement offuel assemblies to a safe position.

In addition, actionmust be immediately initiated to restore boron concentration to within limit.If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3would not specify any action. If moving fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement isindependent of reactor operation.

Therefore, inability tosuspend movement of fuel assemblies is not sufficient reasonto require a reactor shutdown.

SURVEILLANCE SR 3.7.15.1REQUI REMENTSThis SR verifies that the concentration of boron in thespent fuel pool is within the required limit. As long asthis SR is met, the analyzed incidents are fully addressed.

The Surveillance Frequency is controlled under theSurveillance Frequency Control Program.REFERENCES

1. UFSAR, Section 9.1.2.2. PVNGS Operating License Amendments 82, 69 and 54 forUnits 1, 2 and 3, respectively, and associated NRCSafety Evaluation dated September 30, 1994.3. 13-N-001-1900-1221-1, "Palo Verde Spent Fuel PoolCriticality Analysis,"

ABB calculation A-PV-FE-0106, revision 3, dated January 15, 1999.PALO VERDE UNITS 1,2,3B37152RVSO

.B 3.7.15-2REVISION Ifer FP ransition Fuel Storage Pool Boron Concentration B 3.7.15B 3.7 PLANT SYSTEMSB 3.7.15 Fuel Storage Pool Boron Concentration

,an(time.BASESBACKGROUND As dscrbedin.CO 3.7.17. "Spent Fuel Assembly Afuel decieassembliesi tre stored in the spent fuel racks accordance with ,4iteria based on initial enrichment

~discharge burnup' Although the .ater .pent fue pool is .. nomly borte to 2l50 ppm the critei that limtthe strg of.. a fue to sp.cifi rac location-s, isoluble boron concentration of m is required tomaintain the spent fuel pool kef _< 0.ming the mostlimiting fuel mishandling accident. 1460APPLICABLE SAFETY ANALYSESThere could also bea misload ofmultiple fuelassemblies into fuelrack locations notallowed by LCO3.7.17.A fuel assembly could be inadvertently loaded into a spentfuel rack location not allowed by LCO 3.7.17 (e.g., anunirradiated fuel assembly or an insufficiently depletedfuel Another type of postulated accident is a fuel assembly that is dropped onto thefully 1 ~ted fuel pool storage rack or between a rack andthe 4mol walls. These incidents could have a positiveactivity effect, decreasing the margin to criticality.

However, the negative reactivity effect of the soluble boroncompensates for the increased reactivity caused by thesepostulated accident scenarios.

The concentration of dissolved boron in the fuel poolsatisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCOThe specified concentration of dissolved boron in the fuelpool preserves the assumptions used in the analyses of thepotential accident scenarios described above. Thisconcentration of dissolved boron is the minimum requiredconcentration for fuel assembly storage and movement withinthe fuel pool.APPLICABILITY This LCO applies whenever any fuel assembly is stored inthe spent fuel pool in order to comply with theTS 4.3.1.1.c design requirement that keff 0.95.(conti nued)PALO VERDE UNITS 1,2,3B37151RVSO B 3.7.15-1REVISION IAfter SFP transition Fuel Storage Pool Boron Concentration B 3.7.15BASES (continued)

ACTIONSA.1 and A.2The Required Actions are modified by a Note indicating thatLCO 3.0.3 does not apply.When the concentration of boron in the spent fuel pool isless than required, immediate action must be taken topreclude an accident from happening or to mitigate theconsequences of an accident in progress.

This is mostefficiently achieved by immediately suspending the movementof fuel assemblies.

This does not preclude the movement offuel assemblies to a safe position.

In addition, actionmust be immediately initiated to restore boron concentration to within limit.If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3would not specify any action. If moving fuel assemblies while in MODE 1, 2, 3. or 4, the fuel movement isindependent of reactor operation.

Therefore, inability tosuspend movement of fuel assemblies is not sufficient reasonto require a reactor shutdown.

SURVEILLANCE SR 3.7.15.1REQUIREMENTS This SR verifies that the concentration of boron in thespent fuel pool is within the required limit. As long asthis SR is met, the analyzed incidents are fully addressed.

The Surveillance Frequency is controlled under theSurveillance Frequency Control Program.REFERENCES

1. UFSAR, Section 9.1.2.2. PVNGS Opcr,:ting ic-c-nse Amcndmcnts

82. 69 3nd{ 51 fo-'r""Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3" (Proprietary),

WCAP-1 8030-P, Revision 0,September 2015.PALO VERDE UNITS 1,2,3B 3.7.15-2REVISION I Before SFP transition ISpent Fuel Assembly StorageB 3.7.17B 3.7 PLANT SYSTEMSB 3.7.17 Spent Fuel Assembly StorageBAS ESBAGCKGROUND The spent fuel storage is designed to store either new(nonirradiated) nuclear fuel assemblies, or burned(irradiated) fuel assemblies in a vertical configuration underwater.

The storage pool was originally designed to storeup to 1329 fuel assemblies in a borated fuel storage mode.The current storage configuration, which allows credit to betaken for boron concentrati.on, burnup, and decay time, anddoes not require neutron absorbing (boraflex) storage cans,provides for a maximum storage of 1209 fuel assemblies in afour-region configuration.

The design basis of the spent fuelcooling system, however, is to provide adequate cooling to thespent fuel during all operating conditions (including fullcore offload) for only 1205 fuel assemblies (UFSAR section9.1.3). Therefore, an additional four spaces are mechanically blocked to limit the maximum number of fuel assemblies thatmay be stored in the spent fuel storage pool to 1205.Region 1 is comprised of two 9x8 storage racks and one 12x8storage rack. Cell blocking devices are placed in every otherstorage cell location in Region 1 to maintain a two-out-of-four checkerboard configuration.

These cell blocking devicesprevent inadvertent insertion of a fuel assembly into a cellthat is not allowed to contain a fuel assembly.

Region 3 is comprised of three 9x8 storage racks and one 9x9storage rack in Units 2 and 3. Region 3 is comprised of four9x8 storage racks and one 9x9 storage rack in Unit 1. Sincefuel assemblies may be stored in every Region 3 cell location, no cell blocking devices are installed in Region 3.Regions 2 and 4 are mixed and are comprised of seven 9x8storage racks and three 12x8 storage racks in Units 2 and 3,Regions 2 and 4 are mixed and are comprised of six 9x8 storageracks and three 12x8 storage racks in Unit 1. Regions 2 and 4are mixed in a repeating 3x4 storage pattern in which two-out-of-twelve cell locations are designated Region 2 and ten-out-of-twelve cell locations are designated Region 4 (see UFSARFigures 9.1-7 and 9.1-7A).

Since fuel assemblies may bestored in every Region 2 and Region 4 cell location, no cellblocking devices are installed in Region 2 and Region 4.(conti nued)PALO VERDE UNITS 1,2,3B371-1RVSO B 3.7.17-iREVISION

[Before SFP transition Spent Fuel Assembly StorageB 3.7.17BASESBACKGROUND (conti nued)The spent fuel storage cells are installed in parallel rowswith a nominal center-to-center spacing of 9.5 inches. Thisspacing, a minimum soluble boron concentration of 900 ppm,and the storage of fuel in the appropriate region based onassembly burnup in accordance with TS Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3 is sufficient to maintain a keff of0.95 for fuel of original maximum radially averagedenrichment of up to 4.80%.APPLICABLE SAFETY ANALYSESThe spent fuel storage pool is designed for non-criticality by use of adequate

spacing, credit for boronconcentration, and the storage of fuel in the appropriate region based on assembly burnup in accordance withTS Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3.

The designrequirements related to criticality (TS 4.3.1.1) arekeff < 1.0 assuming no credit for boron and keff 0.95taking credit for soluble boron. The burnup versusenrichment requirements (TS Figures 3.7.17-1, 3.7.17-2, and3.7.17-3) are developed assuming keff < 1.0 with no credittaken for soluble boron, and that keff 0.95 assuming asoluble boron concentration of 900 ppm and the mostlimiting single fuel mishandling accident.

The analysis of the reactivity effects of fuel storage inthe spent fuel storage racks was performed by ABB-Combustion Engineering (CE) using the three-dimensional Monte Carlo codeKENO-VA with the updated 44 group ENDF/B-5 neutron crosssection library.

The KENO code has been previously used byCE for the analysis of fuel rack reactivity and have beenbenchmarked against results from numerous criticalexperiments.

These experiments simulate the PVNGS fuelstorage racks as realistically as possible with respect toparameters important to reactivity such as enrichment andassembly spacing.The modeling of Regions 2, 3, and 4 included severalconservative assumptions.

These assumptions neglected thereactivity effects of poison shims in the assemblies andstructural grids. These assumptions tend to increase thecalculated effective multiplication factor (keff) of theracks. The stored fuel assemblies were modeled as CE 16x16assemblies with a nominal pitch of 0.5065 inches between fuelrods, a fuel pellet diameter of 0.3255 inches, and a UO(2)density of 10.31 g/cc.(conti nued)PALO VERDE UNITS 1,2,3B37172RVSO 3B 3.7.17-2REVISION 3

IBefore SFP transition]

Spent Fuel Assembly StorageB 3.7.17BASESAPPLICABLE SAFETY ANALYSES(conti nued)KENO-Va calculations were used to construct curves of burnupversus initial enrichment for decay times in 5 yearincrements from 0 to 20 years for both Regions 3 and 4(TS Figures 3.7.17-2 and 3.7.11-3) such that all points onthe curves produce a keff value (including all biases anduncertainties) of < 1.0 for unborated water. Biasesassociated with methodology and water temperature wereincluded, and uncertainties associated with methodology, KENO-Va calculation, fuel enrichment, fuel rack pitch, fuelrack and L-insert thickness, pellet stack density, andasymmetric fuel assembly loading were included.

KENO-Vacalculations were also performed to determine the solubleboron concentration required to maintain the spent fuel poolkeff (including all biases and uncertainties) 0.95 at a95% probability/95%

confidence level. A soluble boronconcentration of 900 ppm is required to assure that the spentfuel pool keff remains 0.95 at all times. This solubleboron concentration accounts for the positive reactivity effects of the most limiting single fuel mishandling eventand uncertainties associated with fuel assembly reactivity and burnup. This method of reactivity equivalencing has beenaccepted by the NRC (Reference

3) and used for numerous otherspent fuel storage pools that take credit for burnup, decaytime, and soluble boron.Most abnormal storage conditions will not result in anincrease in the keff of the racks. However, it is possibleto postulate events, with a burnup and enrichment combination outside of the acceptable area in TS Figure 3.1.17-1, or witha burnup, decay time, and enrichment combination outside ofthe acceptable area in IS Figures 3.7.17-2 or 3.7.17-3, whichcould lead to an increase in reactivity.

These events wouldinclude an assembly drop on top of a rack or between a rackand the pool walls, or the misloading of an assembly.

Forsuch events, partial credit may be taken for the solubleboron in the spent fuel pool water to ensure protection against a criticality accident since the staff does notrequire the assumption of two unlikely, independent, concurrent events (double contingency principle).

Although asoluble boron concentration of only 900 ppm is required toassure that keff remains 0.95 assuming the single mostlimiting fuel mishandling event, TS 3.7.15 conservatively requires the presence of 2150 ppm of soluble boron in thespent fuel pool water. As such, the reduction in keff causedby the required soluble boron concentration more than offsetsthe reactivity addition caused by credible accidents, and thestaff criterion of keff 0.95 is met at all times.(conti nued)PALO VERDE UNITS 1,2,3B37.-3RVSO 3B 3.7.17-3REVISION 3

IBefore SEP transitionI Spent Fuel Assembly StorageB 3.7.17BASESAPPLICABLE The criticality aspects of the spent fuel pool meet theSAFETY ANALYSES requirements of General Design Criterion 62 for the(continued) prevention of criticality in fuel storage and handling.

The spent fuel pool heat load calculations were based on afull pool with 1205 fuel assemblies.

From the spent fuelpool criticality

analysis, the number of fuel assemblies that can be stored in the four-region configuration is1209 fuel assemblies.

The design basis of the spent fuelcooling system, however, is to provide adequate cooling tothe spent fuel during all operating conditions (including full core offload) for only 1205 fuel assemblies (UFSAR section 9.1.3). Therefore, an additional four spacesare mechanically blocked to limit the maximum number of fuelassemblies that may be stored in the spent fuel storage poolto 1205.The original licensing basis for the spent fuel pool allowedfor spent fuel to be loaded in either a 4x4 array or acheckerboard array, depending on the use of borated poison.A fuel handling accident was assumed to occur with maximumloading of the pool. The fuel pool rack construction precludes more than one assembly from being impacted in afuel handling accident.

The UFSAR analysis conclusion regarding the worst scenario for a dropped assembly (in whichthe horizontal impact of a fuel assembly on top of the Spentfuel assembly damages fuel rods in the dropped assembly butdoes not impact fuel in the stored assemblies) continues tobe limiting.

The spent fuel assembly storage satisfies Criterion 2 of 10CFR 50.36 (c)(2)(ii).

LCO The restrictions on the placement of fuel assemblies withinthe spent fuel pool, according to Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3 in the accompanying LCO, ensures that the keff ofthe spent fuel pool will always remain < 1.0 assuming thepool to be flooded with unborated water. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool according to Figures 3.1.17-1, 3.1.17-2, and 3.7.17-3 in the accompanying LCO.Specification 4.3.1.1 provides additional details for fuelstorage in each of the four Regions.(conti nued)PALO VERDE UNITS 1,2,3B371-4RVSO B 3.7.17-4REVISION IBefore SFP transition Spent Fuel Assembly StorageB 3.7.17BASESAPPLICABILITY This LCO applies whenever any fuel assembly is stored in thespent'fuel pool.ACTIONS A. 1Requi red Acti on A.1 i is modi fi ed by a Note i ndi cating thatLCO 3.0.3 does not apply.When the configuration of fuel assemblies stored in the spentfuel pool is not in accordance with Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3, immediate action must be taken tomake the necessary fuel assembly movement(s) to bring theconfiguration into compliance with Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3.

If moving irradiated fuel assemblies while in MODE 5 or 6,LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuelmovement is independent of reactor operation.

Therefore, ineither case, inability to move fuel assemblies is notsufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.17.1REQU IREMENTSThis SR verifies by administrative means that the initialenrichment and burnup of the fuel assembly is in accordance with Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3 in theaccompanying LCO and Speci fi cati on 4.3.1.1.To manually determine the allowed SFP region for a fuelassembly, the actual burnup is compared to the burnuprequirement for the given initial enrichment andappropriate decay time from Figure 3.7.17-1, 3.7.17-2, or3.7.17-3.

If the actual burnup is greater than or equal tothe burnup requirement, then the fuel assembly is eligibleto be stored in the corresponding region. If the actualburnup is less than the burnup requirement, then thecomparison needs to be repeated using another curve for alower numbered region. Note the fol lowing :(continued)

PALO VERDE UNITS 1,2,3B371-5RVSO 3B 3.7.17-5REVISION 3

SBefore SFP transition ISpent Fuel Assembly StorageB 3.7.17BASESSURVEILLANCE

• that a fuel assembly that does not meet the burnupREQUIREMENTS requirement for Region, 2 must be stored in Region 1,(conti nued)* that any fuel assembly may be stored in Region 1,* that any fuel assembly may be stored in a lower numberedregion than the region for which it qualifies becauseburnup requirements decrease as region numbers decrease(refer also to Tech Spec 4.3.1.1),

• and that comparing actual burnup to the burnuprequirement for zero decay time will always be corrector conservative.

REFERENCES

1. UFSAR, Sections 9.1.2 and 9.1.3.2. PVNGS Operating License Amendments 82, 69, and 54 forUnits 1, 2, and 3 respectively, and associated NRCSafety Evaluation, dated September 30, 1994.3. Letter to T. E. Collins, U.S. NRC to T. Greene, WOG,"Acceptance for Referencing of Licensing TopicalReport WCAP-14416-P, Westinghouse Spent Fuel RackMethodology (TAC NO. M93254)",

October 25, 1996.4. 13-N-001-1900-1221-1, "Palo Verde Spent Fuel PoolCriticality Analysis,"

ABB calculation A-PV-FE-0106, revision 03, dated January 15, 1999.5. Westinghouse letter NF-APS-10-19, "Criticality SafetyEvaluation of the Spent Fuel Pool Map with a ProposedRegion 3 Increase,"

dated February 25, 2010.PALO VERDE UNITS 1,2,3 B371- EIIN~B 3.7.17-6REVISION IAfter SFP transition Spent Fuel Assembly StorageB 3.7.17* The design basis of the spent fuel pool cooling system is toB 3.7 PLANT SYSTEMS IProvide adequate cooling to the spent fuel pool during allIoperating conditions (including full core offload) for up to 1205B 3.7.17 Spent Fuel Assembl lfuel assemblies (UFSAR Section 9.1.3).BASES /I l tBACKGROUND The spent fuel is designed ostor2/'2h4 new(nonirradiated) fuel asse ~lies,--g¢ burned(irradiated) fuel assemblies in Jvertical configuration underwater.

The storage pool sorgnally designed to storeup to 1329 fuel assemblies i eae ue trg oeThe current ,storg config..

ration..

w4. ..hich allow credit. betaken for boron.conentration burnup,..

an dea time, and..dones mnot requirelm mneutron absorbing (boraflex)v ca-nsprovides-ltc a maximu~m stor=ag of 1209Q "fuel assemblies¢ in afour reio configuration=.

.The desig basis of the spent fue,,lcooling syste,-,

however, is to provide adequatea cooling1 to thecore offload frl on nly 1205 asseamblie (UFSTAR sct,,'io~n a 1-- -'- Therefore, an =additional four.spaces.are.mechanically..

bmlockedn to limitf t-he mai~m~um nulmberA t-hatCell de,,'ie are..pnlaceda in ever... otherstoragea cell location in Region 1 to mainti-mn a out- offour checkerbar configuration.

blockin devicespreventf inadverte4-nt-insertion ofc a fueal into a cellthat i not ..alloe to contain a fuel assembly.

sto-rr~age r~ack- in UI I-4t- 2 and'- 3. Regio n 3 is co .v-mprise of,- fou"r,,,9x3 sto"rage,- and oneh 9v9 stor,-age,,-

r,.ack. in- Unit- 1. 4Refue~l asse. mblies- may be kstoredw in evry,,n, Regio"-n 3 ,.-c ll loation4,'

no cell blocking devices are installed in Region 3.Regions 2 andi A are mived- rand4 are seve\-n 9v8 ra-,ck and three 12x r~acks,,

in Un-it -2c and 3,Regio~ns 2 and I nar m'ixed and are of sixv 9v8 storagare mixed, in a repatnga-n 3vA storwage attern in w.hic-h twor outof cell locations are designated Region-,,

2 and ten outof twelve! cell. locations

..... designated Region I (see-, UFS^RFigres 1 "7 and 9.1 7A). S1nc ..-fuel asse-,mblies may, bestored in e,.ery Region 2 and Region '1 cell location no.. celblocking de,,ice ..... installed in Region 2 =an Region Isecontinued PALO VERDE UNITS 1,2,3B3.171RVSO 4B 3.7.17-1REVISION Insert I for TS Bases 3.7.17 page B 3.7.17-1The spent fuel storage cells are installed in parallel rows with a nominal center-to-center spacing of 9.5 inches. This spacing, a minimum soluble boron concentration of 1460 ppm, theuse of neutron-absorbing panels, and the storage of fuel in the appropriate region based on fuelassembly initial enrichment, discharge burnup, and decay time in accordance with TS Tables3.7.17-1 through 3.7.17-5 is sufficient to maintain keff< 0.95 for fuel of initial maximum radiallyaveraged enrichment of up to 4.65 wt%.Disused CEAs, in-core instruments, and other material is stored in trash cans. A trash can maybe stored in any location that is approved to store a fuel assembly.

No special nuclear material(SNM) may be stored in a trash can.

IAfter SFP transitionI Spent Fuel Assembly StorageB 3.7.17BASESBAAC"KG RUDf/ KF(conti4nued)4 The, fucl storage cclls installed4 in parallel rows,wit4h a nominal cntne-r to nnl cnter-spacing of 9. innHch. ThLspacingn a

slubklc, boron co'-nceant-ration of ppmassembkl,,

b 'urnp i/n accordance4

... th. 4 -.T--

3.17 1,_3.7.17 2, andl 3.T.1T 3 is sufficie'-4nt-to maintain keTT-of0.9 foK r, fueJl ofz orgna,, nl maxi{mum radia,4:lly.

ofi upn to- /1.0/.APPLICABLE Thc ...nt fu-el strge....

pol is dest"igned,4 for' nonrlSAFETY ANALYSES critcalty' by ' "se of

spacing, credi.t-for. boron and3'1 tlhe "fueali4n theappn,'nropriate-TS: FZ{TiguresC 3.7.17 1, 3.7.17 2 3,4 an 37117 3. Th& -.de-l-gmke, i 1.0 assu-mi ng no credi t for boron and _4 Q--_c_taki"ng credit- fo sonclulehl boron. The burnup The analysis of-effe'tII'ct of fuel_ storage9 inInsrt 1th fue,,l soag k-vr acks whas pefome by BB :Combust/ion Engineering.-

(CE)usin the threen,44n dime.'-,.,-nsinlMot-arocd hEO A with th-e updated #an-1f~

ENDF/ neutron:-l cros'/Insert section.

library.....Th E.....code ha been....

previously..used.by L I~~~C for the, analysis of~n fuel 4-rack' rea~ctivity~ haver ben.,,-,.,,

bencmark\I d agaih l-ns resu:3lts, fAn-,rom numerousR c: nritical storagem, ra:lck as as possblel w.,ith respectn4 toThe of DRegio~ns

2. :3,nd4 '1 inc'ludeda

seeal, neglecPted4 theTheeassumptoions',n

'tend4 to increnase tlhecalculated3-n, e,-ffective,,,

multlication-:34,-,,

facto',r-,

(keT of t',-heSThe stored fuel assembli!es w.ere modeled as CE 1..x16vassembl'ies writh a3 nomilnal p'itc-h of 0.5065 "inches betweean fuealrods,,4 a fuerl pellet-of 0.325 inhe,,",ac nd a3 llO2)density of 10.31 g/cc.(conInIued) lbUPALO VERDE UNITS 1,2,3B371-2RVSO B 3.7.17-2REVISION

%

Insert 2 for TS Bases 3.7.17 page B 3.7.17-2The nuclear criticality safety analysis in References 1 and 2 considered the following reactivity-increasing accidents:

• Misload of a single assembly into an unacceptable storage location* Multiple assemblies misloaded in series due to a common cause* Spent fuel pooi temperature outside the allowable operating range* Dropped and misplaced fresh fuel assembly* Seismic event* Inadvertent removal of a NETCO-SNAP-IN~

rack insertIn each case, the spent fuel assembly storage met the requirements of 10 CFR 50.68(b)(4).

Thus, the spent fuel storage facility is designed for noncriticality by use of adequate spacing,and neutron absorbing panels considering initial enrichment, fuel burnup, and decay time.

IAfter SEP transitionI Spent Fuel Assembly StorageB 3.7.17BASE[SAPPLICTfABLEI

,AI'E'TV

,ANAILVSES versus. initial1 enrichmen fo dccay tims in 5 ..e.rincrements from 0 to 20 years for both Regions 3 and 4(TS 3.7.17 2 and 3.7.17 3) uciih that- all poimnt-s ornt'he curves a kr valu(inc/mludingr all bi.ases .andu-ncertainties) of 1.0 for.. unboratcd wae...assccriated,-

wit4h met-hodo"log nd wm- .ater, tenmpr~n3-atre wereinluedIcl...

and unerante assuciaedtainthiethodologyat QKEO nr',a calculation fuel, enrichme, nt,l fue rckl pitch, fuel.-,

requret maintainthespenful pool1 ....he1,, (Incui ngm'-4m' allC biasesP un'ertaF'inties) 4h~

95 r, obabiCll ty/95,%PIP confidene-h.-

leve.3P ,--A,,4!

soluble, boromn Hco4ncentratmiokn of,900ppm i eurdt sueta h pnMost' abnr,,,,mal

storage, will not-H resuclt in anincreasea in the ke, of te racks.~ Howevr.,

, it- is possibletlhe aceptab-3le area in TS 3.7.17 1.',, orwitht-he accepntable area in TS Figuresc 3.7.17 2 or 3.7.17 3, w.hichcould latonincrease inreactivity.

These evnt wulincluden an asse.,,mbly.

dropr-topof a rac orI a rack,the pool walls, or the misloading of a~n assembly.

Foboron.in the spent fuel poo water to ensure..protection requ~l'ire tlhe of'- tl-wo' unlikenly, indepndnt,~mn solubl boron concentratio of only 900 ppm is require tomassure. that h, remain 4o 0.9 assuming.

t£. he single. mostfuel poom'l water.P As schl," t-he reductiofnm in h,, Pcauseby%1 required4 sof' lub~le boroPn concentration'{'

t'han staff criterion of ke, _ 0.9 is met at all times.PALO VERDE UNITS 1,2,3B37.73RVSO B 3.7.17-3REVISION

%

IAfter SFP transitionI Spent Fuel Assembly StorageB 3.7.17BASESADDI The criticlity a.pect. of t he spent fuel pool meet theThe spent poo,, heat load, calculations w;ere based on afu, ll pool 1205F fuenl assembl-ies.

Frm t-rhe spent fueilpool0+',

criticalt nalsiThem Mnumber~-

of fueln assemblies, thatl can beC in~ thel~ four- reginl =configuatin i1209Efue asse--imblies.

basi oMf thel spntr fuelare mehnhmica=lly blockedH to -imait the mav-imum number oxf fuenlassemblies t-hat- mayl be3 sltoreda

-in the spntl fueil poollomading of the pool. The fuel pool rac~k coinsrctmlion than. oine assmbl fromA being, impacted uionaful andrling ah .c¥cieant, Th UFSAR analysnis 3cncnml u -ion .hregardn g worst-f scnai for, a dropped assembly (il-n whch-,nifuenl assembly, damages fuel rodsc in the dropped~,-

assembly, not -imnrImr"'

fuel "in tlhe st-ored assemhliesncontinues,n toThe spent fuel assembly storage satisfies Criterion 2 of 10CFR 50.36 (c)(2)(ii).

I abes3..1-Itheog spent7- fulpoa awyeand Fi1u0e3assuming the.LCOare rsictonsitn woh the crtcli~~ety safetye anlslies performefrthe spent f el pool, according to Figure 3.7.....

1.....and 3..17 3in in theaccmpanying LCO,.nue ht h efoSpecification 4.3.1.1 provides additional details for fuelstorage in each of the Regios(conti nued)PALO VERDE UNITS 1,2,3 B 3.7.17-4 REVISION IAfter SFP transition Spent Fuel Assembly StorageB 3.7.17BASESAPPLICABILITY This LCO applies whenever any fuel assembly is stored in thespent fuel pooi.ACTIONSA.1I Tables 3.7.17-1 through 3.7.17-5 and Figure 3.7.17-1'-7I v wRequired Action A.1 is modified by a Note indicating tLCO 3.0.3 does not apply. /When the configuration of fuel assemblies s ed the spentfuel pool is not in accordance with Fgrs 3.7.17 2, and 3.7.17 3, immediate action must; taken tomake the necessary fuel assembly movement(s) bring theconfiguration into compliance with N7!3.7.17-2, and 3.7.17 3.If moving irradiated fuel assemblies while in MODE 5 or 6,LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1. 2. 3. or 4, the fuelmovement is independent of reactor operation.

Therefore, ineither case, inability to move fuel assemblies is notsufficient reason to require a reactor shutdown.

SURVEILLANCE REQU IREMENTSThis SR administrative means that the initialenrichment of the fuel assembly is in accordance with 3 71!7 1 3_717 2. and 3_7.7 3 in theofTbe371-thurough 3..1-5 performanc ofh thi SR.....

-ensure compliance with Specification 4.3.1.1.

l(ot udPALO VERDE UNITS 1,2,3B371-5RVSO B 3.7.17-5REVISION

%

IAfter SFP transitionI Spent Fuel Assembly StorageB 3.7.17BASES requ~lirement-fo~r 2 must e'-h sltored a-in Regnion 1,(continRued)

• that anyJ fuel assembly/

may/ be stored a-in aRegion 1,4-tht- any\ "fuel assembly, may be stored -in a loer, nuimbereda regionn t-han the rego w: .h-ic-h it- qalifies bc au-se~~~~REFERENE

.USR aSo oTc peci0n 9 1.3 n .1.3),* n thatI_ act al bur ,nup '- to the burnup -,A+,"reurmn o eodcytiewllasb orc4.- 13g_ N 001190 1221 1, "Palo,,"

\/erdeH Spent- Fuel Pool/Cr'4-it,-1icality Analysi,c

" ,ABBclulto PV FE/ 0106,Evaluatio-4n ofc the Spntn4 Fuel Pool Map w.,it-h a Proposedca Reg-,ionr 3 Inrease, ac" datedH Febrary, 25-, 2010.-"Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3" (Proprietary),

WCAP-1 8030-P,Revision 0, September 2015.PALO VERDE UNITS 1,2,3 B 3.7.17-6 REVISION Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 4List of Regulatory Commitments Enclosure Description and Assessment of Proposed License Amendment Attachment 4List of Regulatory Commitments Reuaor omimnDue Date/Milestone 1. APS will implement procedural controls torequire verification that fresh fuelassemblies are not placed face-adjacent to one another before completing a fuelmove.2. The transition to the new SEPconfiguration will be completed in all threeunits in accordance with the Spent FuelPool Transition Plan within two years ofthe NRC approval date of the amendment or by December 31, 2019, whichever islater.Within the requested 90-day implementation period following NRC approval.

Within two years of the NRC approval date ofthe amendment or by December 31, 2019,whichever is later.

Enclosure Attachment 8 contains PROPRIETARY information to be withheld under 10 CFR 2.39010 CFR 50.90Maria L. LacalVice President, Nuclears~ale Regulatory

&OversightVed Nuclear Generating StationP.O. Box 52034Phoenix, AZ 85072/° Mail Station 7605Tel 623.393.6491 102-07149-M LL/TNWNovember 25, 2015U. S. Nuclear Regulatory Commission ATI-N: Document Control DeskWashington, DC 20555-0001

Subject:

Palo Verde Nuclear Generating Station (PVNGS)Units 1, 2, and 3Docket Nos. STN 50-528, 50-529, and 50-530License Amendment Request to Revise Technical Specifications toIncorporate Updated Criticality Safety AnalysisIn accordance with the provisions of Section 50.90 of Title 10 of the Code of FederalRegulations (10 CFR), Arizona Public Service Company (APS) is submitting a request for alicense amendment to revise the Technical Specifications (TS) for Palo Verde NuclearGenerating Station Units 1, 2, and 3. The proposed amendment would modify TSrequirements to incorporate the results of an updated criticality safety analysis for both newand spent fuel storage.The enclosure to this letter provides a description and assessment of the proposed changesincluding a technical evaluation, a regulatory evaluation, a significant hazards consideration, and an environmental consideration.

The enclosure also contains eight attachments.

Attachment 1 provides the marked-up existing TS pages. Attachment 2 provides the revised(clean) TS pages. Attachment 3 provides the marked-up TS Bases pages to show theproposed changes.This submittal contains new regulatory commitments (as defined by NEI 99-04, Guidelines for Managing NRC Commitment

Changes, Revision

0) to be implemented, which areidentified in Attachment

4. Attachment 5 provides a non-proprietary version of the criticality safety analysis.

Attachment 6 provides a material qualification report for NETCO-SNAP-IN neutron absorbing spent fuel pool rack inserts.Attachment 7 is an affidavit signed by Westinghouse Electric Company LLC that sets forththe basis on which the proprietary information in Attachment 8 may be withheld from publicdisclosure by the Commission and addresses with specificity the considerations listed in 10CFR 2.390(b)(4).

Correspondence with respect to the proprietary aspects of Attachment 8A member of the STAR!S (Strategic Teaming and Resource Sharing)

Alliance Callaway

• Diablo Canyon " Palo Verde

• Wolf CreekAttachment 8 transmitted herewith contains PROPRIETARY information.

When separated from Attachment 8, this transmittal document is decontrolled.

102-07149-M LL/TNWA-TEN: Document Control DeskU. S. Nuclear Regulatory Commission LAR to Incorporate Updated Criticality Safety Analysis in TSPage 2or the supporting Westinghouse affidavit should reference Westinghouse letter numberCAW-15-4271 and be addressed to James A. Gresham,

Manager, Regulatory Compliance, Westinghouse Electric

Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry

Township, Pennsylvania 16066.Attachment 8 is the Criticality Safety Analysis for Palo Verde Nuclear Generating StationUnits 1, 2, and 3, WCAP-18030-P (Proprietary),

which contains information proprietary toWestinghouse Electric Company LLC.A public pre-submittal meeting was held with the NRC on May 11, 2015 (Agency DocumentAccess and Management System [ADAMS] accession number ML15140A314) to discuss thecriticality safety analysis performed in support of this license amendment request.

A follow-up public conference call to address action items from the May 11, 2015, pre-submittal meeting was held on September 1, 2015 (ADAMS accession number ML15286A028).

In accordance with the PVNGS Quality Assurance

Program, the Plant Review Board and theOffsite Safety Review Committee have reviewed and approved the proposed amendment.

By copy of this letter, this license amendment request is being forwarded to the ArizonaRadiation Regulatory Agency in accordance with 10 CFR 50.91(b)(1).

APS requests approval of the proposed license amendment by October 1, 2017, and willimplement the TS amendment within 90 days following NRC approval.

This request isnecessary to complete the Spent Fuel Pool Transition Plan by the end of 2019.Should you have any questions concerning the content of this letter, please contact ThomasWeber, Department Leader, Nuclear Regulatory

Affairs, at (623) 393-5764.

I declare under penalty of perjury that the foregoing is true and correct.Executed on z 2-< .(Date)Sincerely, M LL/TN W/J R/af

Enclosure:

Description and Assessment of Proposed License Amendment cc: M. L. Dapas NRC Region IV Regional Administrator M. M. Watford NRC NRR Project Manager for PVNGSL. J. KIoss NRC NRR Project ManagerC. A. Peabody NRC Senior Resident Inspector for PVNGSA. V. Godwin Arizona Radiation Regulatory'Agency (ARRA)T. Morales Arizona Radiation Regulatory Agency (ARRA)

Enclosure Description and Assessment of Proposed License Amendment TABLE OF CONTENTS1.0 SUMMARY DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Proposed Changes to the Technical Specifications 2.2 Need for Proposed Changes3.0 TECHNICAL EVALUATION 3.1 Spent Fuel Pool Analysis3.2 New Fuel Storage and Fuel Transfer Equipment Analysis3.3 Spent Fuel Pool Transition Plan4.0 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements

4.2 Precedent

4.3 Significant Hazards Consideration

4.4 Conclusion

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

ATTACHMENTS

1. Marked-up Technical Specifications Pages2. Revised Technical Specifications Pages (Clean Copy)3. Marked-up Technical Specifications Bases Pages4. List of Regulatory Commitments

5. Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3(Non-proprietary),

WCAP-1 8030-NP, Revision 0, September 20156. Material Qualification Report of MAXUS for Spent Fuel Storage, NET-300047-07 Rev 1,November 20157. Westinghouse Application for Withholding Proprietary Information from PublicDisclosure, CAW-15-4271, September 3, 20158. Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3(Proprietary),

WCAP-18030-P, Revision 0, September 2015 LIST OF ACRONYMSANP AREVA PVNGS Lead Test Assembly Combustion Engineering 16x16 FuelAPS Arizona Public Service CompanyENDF Evaluated Nuclear Data FileFHE Fuel Handling Equipment IFSR Intermediate Fuel Storage RackLAR License Amendment RequestLER Licensee Event ReportNFS New Fuel StorageNGF Combustion Engineering 16x16 Next Generation FuelPVNGS Palo Verde Nuclear Generating StationSFP Spent Fuel PoolSTD Standard Combustion Engineering 16x16 FuelTS Technical Specification(s)

VAP Value Added Pellet Combustion Engineering 16x16 Fuelii Enclosure Description and Assessment of Proposed License Amendment 1.0 SUMMARY DESCRIPTION The proposed amendment would revise Palo Verde Nuclear Generating Station (PVNGS)Renewed Operating License Nos. NPF-41, NPF-51, and NPF-74 to amend the Technical Specifications (TS) to incorporate the results of the updated criticality safety analysis, WCAP-1 8030-P (Reference 6.1). The proposed amendment will correct a non-conservative TSregarding NRC approved License Amendment Number 125, which describes the currentlicensing bases for the criticality safety analysis for PVNGS. This is further discussed in Section2.2 of this proposed amendment.

This enclosure provides a description and assessment of the proposed changes including atechnical evaluation, a regulatory evaluation, a significant hazards consideration, and anenvironmental consideration.

The enclosure also contains eight attachments.

Attachment 1provides the marked-up existing TS pages. Attachment 2 provides the revised (clean) TSpages. Attachment 3 provides the marked-up TS Bases pages to show the proposed changes.This submittal contains new regulatory commitments (as defined by NEI 99-04, Guidelines forManaging NRC Commitment

Changes, Revision

0) to be implemented, which are identified inAttachment

4. Attachment 5 provides a non-proprietary version of the criticality safety analysis.

Attachment 6 provides a material qualification report for NETCO-SNAP-IN neutron absorbing spent fuel pool rack inserts.Attachment 7 is an affidavit signed by Westinghouse Electric Company LLC that sets forth thebasis on which the proprietary information in Attachment 8 may be withheld from publicdisclosure by the Commission and addresses with specificity the considerations listed in 10CFR 2.390(b)(4).

Attachment 8 is the Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units1, 2, and 3, WCAP-1 8030-P (Proprietary),

which contains information proprietary toWestinghouse Electric Company LLC.2.0 DETAILED DESCRIPTION 2.1 Proposed Changes to the Technical Specifications The following specific TS changes are proposed as part of the updated criticality safety analysis.

Marked-up TS pages are provided in Attachment 1 and revised (clean) TS pages are providedin Attachment 2.TS 3.7.17, Spent Fuel Assembly Storage-Revise the LCO statement and surveillance requirement statement to reflect theupdated spent fuel assembly storage requirements resulting from WCAP-18030-P

-Add Tables 3.7.17-1 through 3.7.17-5 to define the Fuel Regions-Replace Figure 3.7.17-1 to define the Allowable Storage Arrays-Delete Figures 3.7.17-2 and 3.7.17-3 (information is replaced by Tables 3.7.17-1through 3.7.17-5)

I Enclosure Description and Assessment of Proposed License Amendment

• TS 4.3.1, Criticality

-Change TS 4.3.1.1 .a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1.1.c to read "1460 ppm" [parts per million]

instead of "900 ppm"-Change TS 4.3.1.1.e to refer to Fuel Regions 1 -6as shown in Tables 3.7.17-1through 3.7.17-5-Delete TS 4.3.1.1.f through 4.3.1.1.h (Fuel Regions are defined in Tables 3.7.17-1through 3.7.17-5)

-Change TS 4.3.1 .2.a to read "4.65 weight percent" instead of "4.80 weight percent"-Change TS 4.3.1 .2.d to replace "A nominal 17 inch center to center..."

with"A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center..."

• Add new program TS 5.5.21, Spent Fuel Storage Rack Neutron Absorber Monitoring Program (Proposed TS 5.5.20, Risk In formed Completion Time Program, was submitted on July 31, 2015 [Agency Document Access and Management System (ADAMS)accession number ML15218A300])

The TS Bases will also be revised for consistency with the proposed TS changes and withWCAP-1 8030-P. A markup of the TS Bases pages reflecting these changes is provided inAttachment 3 for information.

The proposed TS Bases changes will be implemented inaccordance with TS 5.5.14, Technical Specifications (TS) Bases Control Program, at the sametime that the TS changes in the approved license amendment request (LAR) are implemented.

2.2 Need for Proposed ChangesIn March of 2000, the NRC approved License Amendment Number 125, which describes thecurrent licensing bases for the criticality safety analysis for PVNGS. That amendment increased the storage capacity of the spent fuel pools (SFPs) by allowing credit for soluble boron anddecay time in the criticality safety analysis.

The amendment also increased the maximumradially averaged fuel enrichment from 4.3 weight percent U-235 to 4.8 weight percent U-235.The methodology that was the basis for that amendment was analogous to that developed inWCAP-14416-P-A, Westinghouse Spent Fuel Rack Criticality Analysis Methodology, which wasreviewed and approved by the NRC for use [NRC Letter, T. E. Collins (NRC) to T. Greene(WOG), "Acceptance for Referencing of Licensing Topical Report WCAP-14416-P, Westinghouse Spent Fuel Rack Methodology (TAC No. M93254),"

dated October 25, 1996]. In2001 and 2004, Arizona Public Service Company (APS) submitted license amendment requests(LARs) to the NRC that would support replacement of the steam generators and authorize subsequent operation at an increased maximum power level of 3990 Megawatts thermal (a 2.94percent increase).

The NRC approved the amendments for Unit 2 in 2003 (ADAMS accession number ML032720538) and for Units I and 3 in 2005 (ADAMS accession numberML0531 30275).In May of 2013, APS submitted Licensee Event Report (LER) 2013-001-00 (ADAMS accession number ML13133A002),

which reported that certain impacts to the SFP criticality safety analysisapproved in License Amendment 125 had not been considered by APS during the increase inthe maximum power level to 3990 MWt. One of the corrective actions in the LER was to revisethe SFP criticality safety analysis using updated methodology and input parameters and to2 Enciosure Description and Assessment of Proposed License Amendment submit a LAR to correct the non-conservative TS. The criticality safety analysis methodology included in this LAR is based upon the most recent NRC approved guidance of Staff GuidanceRegarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, DSS-ISG-201 0-01(Reference 6.2) and satisfies the actions stipulated in the PVNGS Corrective Action Programand PVNGS LER 201 3-001-00.

3.0 TECHNICAL EVALUATION This LAR documents an updated criticality safety analysis for the PVNGS SFPs, new fuelstorage (NFS) racks, interim fuel storage rack (IFSR) and fuel handling equipment (FHE).Attachment 8 is the plant-specific Westinghouse WCAP-18030-P, Revision 0, Criticality SafetyAnalysis for Palo Verde Nuclear Generating Station Units 1, 2 and 3. A plant-specific NETCO-SNAP-IN0 material qualification report is also included (Attachment

6) since the LAR credits thepresence of neutron poisons in the NETCO-SNAP-IN neutron absorber inserts.

The main bodyof the LAR includes descriptions and summary evaluations, while the attached WCAP andNETCO report provide additional details on topics that include computer codes, fuel designhistory, depletion

analysis, criticality

analysis, as well as the interface, normal, and accidentconditions for PVNGS.The change to TS 4.3.1 .2.d regarding NFS rack spacing is proposed to more accurately reflectthe NFS as-built drawings and the existing NFS criticality safety analysis of record. The text ofTS 4.3.1.2 states "The new fuel storage racks are designed and shall be maintained with...",

which refers to physical dimensions.

Therefore, TS 4.3.1 .2'.d must accurately reflect the as-builtdimensions that must be maintained throughout the life of the plant. The racks have a nominal18-inch center-to-center pitch on the short axis (east-west) and a 31-inch center-to-center pitchon the long axis (north-south).

According to both the existing NFS and updated criticality safetyanalyses, this rack design maintains keff <0.95 during all normal and accident conditions.

3.1 Spent Fuel Pool AnalysisDesign ApproachThe existing SEP storage racks are evaluated for the placement of fuel within the storage arraysdescribed in the proposed TS changes.

Credit is taken for the negative reactivity associated with burnup and post-irradiation cooling time (decay time). Additionally, some SEP storagearrays credit the presence of the neutron poison in the NETCO-SNAP-IN inserts.

Finally, creditis taken for the presence of soluble boron in the SFP for specific conditions.

Compliance for the SFP is demonstrated by establishing limits on the minimum allowable burnup as a function of initial enrichment and decay time for each fuel storage array. Aconservative combination of best estimate and bounding values has been selected to model thefuel in the analysis to ensure that fuel represented by the proposed TS is less reactive than thefuel modeled in the analysis.

Therefore, burnup limits will conservatively bound fuel to be storedin the SFP.3 Enclosure Description and Assessment of Proposed License Amendment Acceptance CriteriaThe objective of the SEP criticality safety analysis is to ensure that the SEP operates within thebounds of 10 CFR 50.68(b)(4):

• If no credit for soluble boron is taken, the keff of the spent fuel storage racks loaded withfuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percentprobability, 95 percent confidence level, if flooded with unborated water.* If credit is taken for soluble boron, the keff of the spent fuel storage racks loaded with fuelof the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percentprobability, 95 percent confidence level, if flooded with borated water, and the keff mustremain below 1 .0 (subcritical),

at a 95 percent probability, 95 percent confidence level, ifflooded with unborated water.Computer CodesThe SEP criticality safety analysis methodology employs the following computer codes andcross-section libraries:

• SCALE 6.1.2 (Reference 6.3) with the 238-group cross-section library based onEvaluated Nuclear Data File (ENDF)/B-VII.

• The two-dimensional transport lattice code PARAGON (Reference 6.4) and its 70-groupcross-section library based on ENDF/B-VI.3 PARAGON is used for simulation of in-reactor fuel assembly depletion to generate isotopics forburnup credit applications in the SEP. PARAGON is generically approved for depletion calculations (Reference 6.4) and has been chosen for this spent fuel criticality safety analysisbecause it has all the attributes needed for burnup credit applications.

There are no SafetyEvaluation Report limitations for the use of PARAGON in UO2 criticality analyses.

Additional discussion of the computer codes is provided in Section 2.3 of Attachment 8.Code Validation ProcessThe validation of the ENDF/B-VII library with the SCALE 6.1.2 CSAS5 module is documented inAppendix A of Attachment

8. The code validation shows that SCALE 6.1.2 is an accurate toolfor calculation of kefffor the applications in this LAR. The benchmark calculations utilize thesame computer platform and cross-section libraries that are used for the design basiscalculations.

3.1.1 Fuel and Fuel Storage Descriptions Four fuel designs were considered for the criticality safety analysis:

• Standard Combustion Engineering 16x16 (STD)* Value Added Pellet Combustion Engineering 16x16 (VAP)* Next Generation Fuel Combustion Engineering 16x16 (NGF)* AREVA PVNGS Lead Test Assembly Combustion Engineering 16x16 (ANP)4 Enclosure Description and Assessment of Proposed License Amendment Section 3.1.2 of Attachment 8 discusses the non-mechanical fuel features which are important to criticality safety and how they impact the number of distinct fuel designs to be considered inthe analysis.

PVNGS has three SFPs, one in each unit, that are identical in layout. Each SFP contains asingle rack design and each SEP is surrounded by a concrete wall with a stainless steel liner.The presence of neither the SEP concrete wall nor liner is credited in the criticality safetyanalysis.

All storage arrays are conservatively assumed to be radially infinite.

The SEP storage racks are made up of individual

modules, each of which is an array of fuelstorage cells. The storage racks are comprised of 17 modules:

twelve 8x9 arrays, four 8x12arrays, and one 9x9 array. The storage racks are stainless steel honeycomb structures withrectangular fuel storage cells compatible with fuel assembly materials and the spent fuelborated water environment.

The fuel assembly spacing of a nominal 9.5 inches center-to-center distance between adjacent storage cell locations is a minimum value after allowances are madefor rack fabrication tolerances and predicted deflections resulting from a safe shutdownearthquake.

The storage racks are designed to maximize the number of storage cells available (minimize storage cell pitch) to be used in a checkerboard pattern of fresh (i.e., new) fuel and emptystorage locations.

Trashcans in the SEP store various non-fissile materials, such as discarded control element assemblies and in-core instrumentation tubes, filters, and reconstitution materials.

The SEP racks contain a stainless steel L-insert in every other cell location as shown in PVNGSUpdated Final Safety Analysis Report (UFSAR) Figure 9.1-5 to help center the new fuelassemblies within this space. The stainless steel L-inserts are offset 11/16 inch from the cellwall. This reduces the positional uncertainty or the eccentric loading positions of fuel in theracks. There is not currently a surveillance program for the stainless steel L-inserts and one willnot be included as part of this submittal for the following reasons:* The industry has a large body of operating experience with stainless steel in SEPenvironments in which nothing suggests the stainless steel L-inserts would becomeincapable of performing their function.

• The stainless steel L-inserts are analogous to the SEP racks in their function and there isno precedent for requiring a surveillance program of the racks themselves.

A NETCO-SNAP-IN rack insert will not be installed in a cell that contains a stainless steelL-insert.

The NETCO-SNAP-IN rack inserts are described in Attachment 6.3.1.2 Depletion AnalysisThe depletion analysis is a vital part of any SEP criticality safety analysis which uses burnupcredit. The isotopic inventory of the fuel as a function of burnup is generated through thedepletion

analysis, thus the inputs used need to be carefully considered.

Section 4 ofAttachment 8 describes the methods used to determine the appropriate inputs for thegeneration of isotopic number densities to conservatively bound fuel depletion and storage.Some of the salient points regarding the approach to the depletion analysis include:5 Enclosure Description and Assessment of Proposed License Amendment

• The isotopic number densities generated by the fuel depletion calculations aredifferentiated by fuel enrichment and decay time after discharge.

The fuel has isotopicnumber densities which are calculated at enrichments of 3.0, 4.0, and 5.0 weight percentU-235 and decay times of 0, 5, 10, 15, and 20 years.* The soluble boron concentration in the reactor during operation impacts the reactivity offuel being discharged to the SEP. Boron is a strong thermal neutron absorber and itspresence hardens the neutron energy spectrum in the core, creating more plutonium.

Itis important to account for the presence of soluble boron during reactor operation toensure this impact is adequately accounted for in the isotopic generation.

• The fuel temperature during operation impacts the reactivity of fuel being discharged tothe SFP. Increasing fuel temperature increases resonance absorption in U-238 due toDoppler broadening which leads to increased plutonium production, increasing thereactivity of the fuel.* The limiting distributed axial burnup profiles are used with the uniform axial burnupprofile to calculate the burnup limits.* The limiting axial moderator temperature profiles are used with axially distributed anduniform burnup profiles to calculate the isotopics used in generating the burnup limits.Selecting an appropriate moderator temperature profile is important as it impacts themoderator density and the neutron spectrum during depletion, as discussed inNUREG/CR-6665 (Reference 6.5). An appropriate moderator temperature ensures theimpact of moderator density on the neutron spectral effects is bounded, conservatively biasing the isotopic inventory of the fuel.* Burnable absorber usage has been considered for the analysis and conservative assumptions have been utilized to bound the effects of burnable absorbers on fuelisotopics.

The burnable absorbers that have been used include both discrete andintegral burnable absorbers.

• The PVNGS fuel management strategy uses radial enrichment zoning to control fuel rodpower peaking.

Individual assemblies may contain two or three different fuel rodenrichments which are used to control peaking factors.

A study was performed todetermine the reactivity impact of operating with radial enrichment zoning instead ofuniform radial zoning.* All four of the different fuel designs listed in Section 3.1.1 and the conditions in whichthose designs were operated, or are planned to be operated, were considered in thedepletion analysis.

It became clear that the NGF design would be limiting throughout life.Therefore, the NGF design was used to develop the isotopics used in the spent fuelreactivity calculations.

• The parameters used in the final depletion calculations include core operation parameters, fuel assembly dimensions, axial burnup profiles, and moderator temperature profiles.

3.1.3 Spent Fuel Pool Criticality AnalysisKENO is the criticality code used to determine the absolute reactivity of burned and fresh fuelassemblies loaded in storage arrays. The dimensions and tolerances of the design basis fuelassembly and the fuel storage racks are the basis for the KENO models used to determine theburnup requirements for each fuel storage array, and to confirm the safe operation of the SEP6 Enclosure Description and Assessment of Proposed License Amendment under normal and accident conditions.

The trashcan characteristics are also modeled in thecriticality analysis.

Differences between fuel types include changes in fuel rod dimensions, such as pellet andcladding dimensions, and structural components, such as grid material and volumes.

Each ofthe fuel types which have been, or are planned to be, operated at the plant were considered.

The bounding fuel assembly design for the analysis has been determined as described inSection 4.3 of Attachment 8.Burnup Limit Generation To ensure safe operation of the PVNGS SFPs, the analysis defines fuel storage arrays whichdictate where assemblies can be placed in the SFPs based on enrichment (weight percent U-235), average burnup (GWd/MTU),

and decay time (years) since discharge.

Each assembly inthe reactor core depletes under slightly different conditions and can have a different reactivity atthe same burnup. This is accounted for in the analysis by using a combination of depletion parameters that together produce a bounding isotopic inventory throughout life. Additionally, while fuel manufacturing is a very tightly controlled

process, assemblies are not identical.

Reactivity margin is added to the KENO reactivity calculations for the generation of burnup limitsto account for manufacturing deviations.

Assembly storage is controlled by defining allowable storage arrays. An array can only bepopulated by assemblies of the fuel region defined in the array definition or a lower reactivity fuel region. Fuel regions are defined by assembly burnup, initial enrichment, and decay time.Reactivity biases are known variations between the real and analyzed system, and theirreactivity impact is added directly to the calculated keff. Uncertainties are random dispersions around a nominal, measured quantity.

Their impact is added to the calculated keff as the squareroot of the sum of the squares of the uncertainties.

The following biases and uncertainties areaccounted for in the analysis.

A detailed discussion of biases and uncertainties is provided inSection 5.2.3 of Attachment 8.* Reactivity effect of manufacturing tolerances

• Burnup measurement uncertainty

• Depletion uncertainty

• Fission product and minor actinide worth bias* An operational uncertainty of 0.002 Ak is an additional conservativism which is added tothe conservatism inherent in the specific power histories from reactor operation

• Eccentric fuel assembly positioning

• Uncertainty in the predictive capability of SCALE 6.1.2 and the associated cross-section library* SFP temperature bias within the allowable operating range* Borated and unborated biases and uncertainties 7

Enclosure Description and Assessment of Proposed License Amendment Interface Modelingq Interfaces are the locations where there is a change in either the storage racks or the storagerequirements of the fuel in question.

At PVNGS, each SFP has a single storage rack design.Therefore, the only interfaces that exist are those between arrays within the single storage rackdesign and the only interface conditions that need to be addressed in the analysis are thosebetween different fuel storage arrays. Additional details are provided in Section 5.3 ofAttachment 8.Normal Conditions Considered in the Criticality Safety AnalysisThere are five major types of normal conditions beyond the storage of fuel assemblies that areaddressed in the criticality safety analysis.

Type 1 conditions involve placement of components in or near the intact fuel assemblies whilenormally stored in the storage racks. This also includes removal and reinsertion of thesecomponents into the fuel when stored in the rack positions using specifically designed tooling.Examples include control element assemblies and guide tube inserts, such as in-coreinstrumentation tubes. The calculation results show that any components designed to beinserted into an assembly may be stored in a fuel assembly guide tube in the SFP.Type 2 conditions involve evolutions where the fuel assembly is removed from the normalstorage rack location for a specific procedure and returned to an allowable cell after completion of the procedure, such as fuel assembly

cleaning, inspection, reconstitution, or sipping.

Theseare bounded by the criticality analysis.

Fuel assembly reconstitution is a normal condition defined as either pulling damaged fuel pinsout of an assembly and reinserting intact pins with less reactivity than the damaged pin, or asremoving undamaged pins from a damaged assembly for insertion in a new assembly.

Damaged pins will be replaced with stainless steel pins or natural uranium pins. Additional information is provided in Section 5.4.2 of Attachment 8.Type 3 conditions involve inserting components that are not intact fuel assemblies into the fuelstorage rack cells. Examples include failed fuel rod baskets and miscellaneous maintenance equipment.

Any components that do not contain fissile materials can replace a fuel assembly ofany fuel region in one of the approved storage configurations.

Type 4 conditions include temporary installation of non-fissile components on the rackperiphery.

Analyses of the storage arrays contained in the criticality analysis assume an infinitearray of storage cells. This assumption bounds the installation of any non-fissile components onthe periphery of racks.Type 5 conditions involve miscellaneous conditions that do not fit into the first four normalcondition types. Examples include usage of fuel handling tools for their intended purpose,miscellaneous debris under the storage racks, and damaged storage cells.Section 5.4 of Attachment 8 provides further details about normal conditions within the SEP.8 Enclosure Description and Assessment of Proposed License Amendment Soluble Boron CreditIn accordance with 10 CFR 50.68, the criticality safety analysis ensures that the maximumcalculated ke., including all biases and uncertainties, meet the kef limit of less than 1.0(subcritical) if flooded with unborated water at a 95 percent probability, 95 percent confidence level. Additionally, the criticality safety analysis demonstrates that if the SFP is flooded withborated water, keff does not exceed 0.95, at a 95 percent probability, 95 percent confidence level.The minimum soluble boron concentration in the SEP to maintain keff < 0.95 for the limitingnormal condition including biases, uncertainties, and administrative margin is 450 ppm. Duringnormal operation, TS 3.7.15 requires a soluble boron concentration of> 2150 ppm, but the SFPboron concentration is maintained between 4000 and 4400 ppm in accordance with Technical Requirements Manual T3. 1.104, Borated Sources -Shutdown, and T3. 1.105, Borated Sources-Operating.

Consideration of Criticality Accidents in the SEPThe following reactivity-increasing accidents are considered and the analysis results areprovided in Section 5.6 of Attachment 8.SAssembly misload into the storage racks -this is the limiting accident which addresses both multiple assemblies being misloaded in series into unacceptable storage locations and the misload of a single assembly into an unacceptable storage location.

A multipleassembly misload is a hypothetical accident where assemblies are misloaded in seriesdue to a common cause. A single assembly misload requires 1100 ppm of boron tomaintain keff -< 0.95. A multiple assembly misload requires 1460 ppm of boron to maintainkef<0.95.* Spent fuel temperature outside operating range -the SEP is to be operated between60°F and 180°F, but under accident conditions this temperature could be higher.* Dropped and misplaced fresh assembly

-the analysis considers the dropping of the fuelassembly from the fuel handling machine during placement of the fuel assemblies in theracks. The dropped assembly could land horizontally on top of the other fuel assemblies in the rack. Additionally, the analysis considers the possibility to misplace a fuelassembly in a location not intended for fuel.* Seismic event -the SEP racks are seismic category I, designed and built to withstand the maximum potential earthquake stresses in this geographic area. Section 5.6.4 ofAttachment 8 provides additional details.o The spent fuel pool racks were originally designed to contain a 188 lb. neutronpoison insert in every cell. These neutron poison inserts were never installed.

Asthe mass of the original design is greater than the mass of the NETCO-SNAP-IN, the original analysis bounds the proposed change.* Inadvertent removal of a NETCO-SNAP-IN rack insert -this is a potential reactivity-increasing accident added by the incorporation of NETCO-SNAP-IN rack inserts.

Theabsence of an insert will cause a reactivity increase due to the loss of neutron absorbing material from the storage array.9 Enclosure Description and Assessment of Proposed License Amendment Fuel used to date at PVNGS has an initial radially averaged enrichment of < 4.55 weightpercent.

Limiting the maximum radially averaged enrichment to 4.65 weight percent mitigates the consequences of a multiple fuel assembly misload event without impacting operational flexibility.

There is no source of water within the fuel building that could reduce the boron concentration ofthe spent fuel pool from the value of 2150 ppm (Technical Specification LCO 3.7.15) to 1460ppm. A fire in the fuel building at elevation 140-ft. is the limiting event for boron dilution and itbounds all normal, seismic, and pipe break scenarios.

The current boron dilution analysisdemonstrates that the limiting boron dilution event, which is fighting a hypothetical fire on the140-ft level of the fuel building, will reduce the boron concentration from the TS limit of 2150ppm to 1900 ppm. This leaves adequate margin to the 1460 ppm credited by the SEP criticality safety analysis.

3.1.4 NETCO-SNAP-IN Rack InsertsThis proposed change would credit NETCO-SNAP-IN rack inserts for criticality control inindividual SEP storage rack cells to ensure that the requirements of TS 3.7.17 and theassociated WCAP-1 8030-P are maintained.

The NETCO-SNAP-IN rack inserts are credited inboth the borated and unborated conditions.

The installation of the NETCO-SNAP-IN rackinserts will be controlled as a design change implemented under the provisions of 10 CFR50.59, Changes, Tests and Experiments, from a structural,

seismic, and thermal-hydraulic perspective.

Attachment 6 describes the NETCO-SNAP-IN rack inserts, including their manufacture, anengineering evaluation, and corrosion testing information.

Neutron Absorber Monitoring Program (TS 5.5.21)Arizona Public Service will institute a performance-based long-term surveillance program for theNETCO-SNAP-IN inserts based on manufacturer recommendations, current industry operating experience, NEI guidance, and NRC safety evaluations for other plants that are using neutronabsorbing inserts.

The long-term surveillance program will evolve as information from thesesources changes and as the data from the PVNGS-specific inspections accumulate.

Thesurveillance program consists of periodic inspections of MAXUS material coupons fromsurveillance assemblies located in the SEPs and periodic inspection of full length inserts.Coupon Inspections Coupons will be selected from MAXUS production

material, identical to the material used tomanufacture the inserts, for periodic inspection.

Individual coupons will be subjected to pre-testand post-test characterizations.

As appropriate for each coupon type, coupon characterizations may include visual inspection, high resolution photography, neutron attenuation, stressrelaxation, blister and pit characterizations, as well as measurement of thickness, length, width,dry weight, and density.A surveillance assembly to which surveillance coupons are attached, also referred to as acoupon tree, will be placed in the related SEP prior to the first installation campaign of NETCO-SNAP-IN inserts and will reside there to support the monitoring program.

Periodically, coupons10 Enclosure Description and Assessment of Proposed License Amendment will be removed and sent to a qualified laboratory for testing.

The coupon trees in the relatedSFP will contain 48 general coupons, 24 galvanic couple coupons, and 24 bend coupons asdescribed below. They will be situated near the center of the active fuel region to maximizeexposure from the surrounding fuel. These coupons will be monitored for changes to theirphysical properties and for changes to their effective areal density or signs of corrosion, whichcould indicate neutron absorber material degradation.

The frequency for coupon removal and inspection is shown in Table 1.Table I -Frequency for Coupon Removal from RacksAfter 10 Years withCoupon Type First Ten Years Acpal efracGeneral 2 coupons every 2 years 2 coupons every 4 yearsBend I coupon every 2 years 1 coupon every 4 yearsGalvanic couples -304L stainless I couple every 6 yearsZircaloy 1 couple every 6 yearsInconel 718 1 couple every 6 yearsGeneral and Galvanic CouponsThe general coupons in each SFP are designed to carry the largest number of performance indicators for the insert material.

They will be subject to pre-examination, post-examination, andacceptance testing in accordance with Table 2.The galvanic couple coupons are composed of a MAXUS material coupon placed in contactwith Zircaloy, Inconel 718, or 304 stainless steel. Eight of each type will be used to produce atotal of 24 galvanic couples per coupon tree assembly in each SEP that will be subject to theinspections listed in Table 2.Bend CouponsOnce installed, the NETCO-SNAP-IN rack inserts assume a constant strain condition within theSEP storage rack cell. This compression leads to internal

stresses, especially at the bend, thatmight make the rack inserts susceptible to stress corrosion cracking.

An examination of theliterature on the subject indicates in general, that high-purity aluminum and low-strength aluminum alloys are not susceptible to stress corrosion cracking.

However, the surveillance bend coupons placed in the related SEP will be maintained under the same strain conditions asthe inserts to provide an indication of unexpected crack phenomena.

These coupons will beheld in capsules that compress them from their initial manufactured bend angle to an angle ofapproximately 90 degrees.

Table 3 provides the inspection details.11 Enclosure Description and Assessment of Proposed License Amendment Over time, the MAXUS material is expected to release some of the strain built up during theinstallation process.

The material has a metal matrix core made from 1000 series aluminum andboron carbide powder with an outer clad made from 5052 aluminum.

Existing literature for 1100series aluminum shows a stress relaxation rate of 58 percent over a period of 20 years. Giventhat 1100 series aluminum is a softer metal than 5052 aluminum, this rate is considered conservative for the MAXUS material due to the 5052 cladding.

The acceptance criterion forstress relaxation is 60 percent over a 20-year period. This rate will be used when determining minimum retention force requirements for the inserts during installation that will still hold theinserts in place during a seismic event after relaxation has occurred.

The bend coupon capsules will be removed from the coupon tree and sent to a qualified laboratory for testing where the coupons will be removed, thus relieving the strain on thecoupons and allowing them to return to an angle greater than 90 degrees.

The change ininternal stress can be correlated to the change in bend angle the coupon forms once it isremoved from the strained condition.

Deviation from the pre-characterized value will determine the amount of stress relaxation over the life of the coupon.The stress relaxation rate is not linear, rather it tends to follow a logarithmic pattern.

Therefore, a more significant loss of stress is expected in the first few years of exposure, but the relaxation rate becomes asymptotic over a longer period of time.*Table 2 -General and Galvanic Coupon Characterizations TetPre- Post- Acceptance I Rejection Characterization Characterization CriteriaVisual (high / Evidence of visualresolution digital indications ofphoto) performance inhibitors.

Dimension

/Min. thickness:

0.005 inch less thannominal thickness (excluding pit locations).

Thickness change: anychange of +0.010 inch I-0.004 inch (excluding pitlocations).

• Length change: anychange of +/- 0.02 inch* Width change: anychange of +/- 0.02 inchDensity /Any change of +/- 5%Areal density g/cm2 Boron-10minimum loading12 Enclosure Description and Assessment of Proposed License Amendment Weight loss as determined by dry change of +/- 5%weightCorrosion rate /< 0.05 mil/yrAnomaly characterization

/** To be determined at thetime of analysis.

• Length and width changes are not applicable for galvanic coupons.** At the presence of anomalies Table 3 -Bend Coupon Characterizations TetPre- Post- Acceptance

/Characterization Characterization Rejection CriteriaVisual (high Evidence of visualresolution indications ofdigital photo) performance inhibitors.

Thickness

' V Min. thickness:

0.005 inch less thannominal thickness (excluding pit locations).

Thickness change: anychange of +0.010 inch /-0.004 inch (excluding pit locations).

Bending V /Change in stress greaterstress than a rate of 60% over20 years *Weight loss as determined by dry change of +/- 5%weightAnomaly characterization

• • To be determined at thetime of analysis.

Stress relaxation rate is not linear. Stress relaxation will be re-evaluated if 60% is exceeded.

• • At the presence of anomalies Full-Length Insert Inspections The combined effects of adequate clearance and infrequent fuel assembly movement willpreclude significant wear of the rack insert. However, to verify NETCO-SNAP-IN materialperformance, a portion of the installed inserts will be subject to in-situ visual inspection andremoval for detailed inspection of wear performance.

For in-situ inspections at the frequency described in Table 4, rack inserts will be visually inspected by camera (while remaining in thestorage racks) to monitor for physical deformities such as bubbling, blistering, corrosion pitting,13 Enclosure Description and Assessment of Proposed License Amendment

cracking, or flaking.

Special attention shall be paid to the development of edge or cornerdefects.A region of high duty spent fuel storage rack cell locations shall be identified for full insertremoval and inspection.

These locations will be monitored for fuel insertion and removal eventsto ensure that their service bounds that of the general population of storage locations.

Onceevery 10 years, an insert will be fully removed from this region and will be inspected inaccordance with Table 5. The thickness measurements at several locations along the full insertlength will be compared with the as-built thickness measurements of the removed insert to verifyit has sustained uniform wear over its service life. A visual inspection of the removed insert willalso be performed.

Table 4 -Frequency for Full Insert Inspections Inspection Type First Ten Years Acpal efracIn-situ 2 inserts every 2 years 2 inserts every 4 yearsRemoval 1 insert every 10 yearsTable 5 -Full Insert Removal Inspection Characterizations TetPre- Post- Acceptance ICharacterization Characterization Rejection CriteriaVisual (high of visualresolution indications ofdigital photo) performance inhibitors.

Thickness thickness:

0.005 inch less thannominal thickness (excluding pitlocations).

Thickness change: anychange of +0.010 inch I-0.004 inch (excluding pit locations).

Retention force Retention force lessthan 50 lbs14 Enclosure Description and Assessment of Proposed License Amendment 3.1.5 Spent Fuel Pool Configuration Control (Human Performance Enhancements)

APS has a multi-tier defense-in-depth program to prevent and mitigate the severity of a scenarioin which multiple fuel assemblies are located in the wrong storage locations.

Specific aspects ofthis program are described below.Control of Move Sheet Generation

• Detailed administrative procedures for the generation of move sheets and revision ofmove sheets* Training and qualification of individuals responsible for generation and revision of movesheets for use and implementation of the new TS proposed in this LAR.* Graphical representation of approved arrays in TS 3.7.17 to minimize the probability ofmisinterpretations

• In accordance with the plant special nuclear materials procedures, APS maintains aRegion Specification Document for each SEP to aid the move sheet preparer and verifierin selecting and verifying proper placement of fuel assemblies.

This document tracks thefollowing for every fuel assembly at PVNGS:-Fuel assembly initial enrichment

-Fuel assembly burnup-Limiting Fuel Region in which the fuel assembly can be stored* Every fuel move is checked against the availability of the space and the eligibility of fuelto be stored there* A member of management confirms that each move sheet was generated in accordance with PVNGS procedures.

There are at least three signatures on each move sheet sent tothe field, including the move sheet preparer, the verifier, and management.

• A move sheet package is a change document that is used to specify and record changesto plant configuration as it relates to special nuclear material.

A move sheet contains asa minimum, the item to be moved, the "from" location, and the "to" location.

Any numberof fuel assemblies can be moved using a single move sheet package.Control of Fuel Movement* The spent fuel handling machine is only operated using approved procedures

• All individuals operating the spent fuel handling machine and acting as independent verifiers are trained in their position, including training on industry operating experience pertaining to fuel misload events* Fuel is moved only as directed by approved move sheets* The correct location of the spent fuel handling machine is independently verified before afuel move begins* The correct location of the spent fuel handing machine is independently verified beforethe fuel is placed in the SEP racks* Continuous communication is maintained between the fuel mover and verifier15 Enclosure Description and Assessment of Proposed License Amendment

• The fuel handlers visually confirm that fresh fuel is not placed in locations that are "face-adjacent" to other fresh fuel assemblies Use of Blockingq DevicesThe assumed limiting misload event at PVNGS involves placing a fuel assembly in a locationthat is required to be empty per TS 3.7.17. APS uses blocking devices to minimize theprobability of a fuel assembly being placed in one of these limiting locations.

Each blockintg device meets the following criteria:

• Physically configured to prevent insertion of a fuel assembly in a fuel storage location* Requires special tools to install or remove the blocking device from a storage location* The tool used to grapple a fuel assembly is physically incapable of grappling a blockingdevice* Designed to preclude falling into a storage location or becoming dislodged during normaloperation

• Will support the full load of a fuel assembly and the fuel assembly grappling tool* Allows continuous water flow through the storage cell0SIs easy to distinguish visually from a fuel assemblyBlocking devices are administratively controlled with the same level of rigor as fuelassemblies

• A blocking device move sheet package is a change document used to specify andrecord changes to plant configuration as it relates to blocking devices.

A blocking devicemove sheet contains, as a minimum, the "from" location and the "to" location of theblocking device. Any number of blocking devices may be moved using a single blockingdevice move sheet package.

Blocking devices will not be moved using the same movesheet package as fuel assemblies.

This restriction prevents a single error from removinga blocking device and placing a fuel assembly in a location that is required to be empty.Confirmation of Configquration ControlIn accordance with existing procedures, a 100 percent serial number check is performed onceper calendar year to ensure that every fuel assembly stored in the SFP matches the SFP maps.This provision limits the amount of time that a misload condition could potentially exist.Mitigqation of a Misload EventIf the controls discussed above are insufficient to prevent a fuel assembly from being misloaded, the following will mitigate the consequences of such an event:* Misload events, including misload events involving multiple fuel assemblies, have beenanalyzed-An adequate soluble boron margin mitigates the misload event. TS 3.7.15requires 2150 ppm of soluble boron* The limiting misload of a single fuel assembly requires 1100 ppm ofsoluble boron to maintain keff <0.9516 Enclosure Description and Assessment of Proposed License Amendment

• The limiting,

credible, multiple misload of placing a fresh fuel assemblyinto every blocked location in the most limiting array (Array C) requires1460 ppm of soluble boron to maintain keff <0.95Placing fresh fuel assemblies face-adjacent to one another is not credible.

A defense-in-depth approach provides

multiple, independent barriers to this event. These barriersinclude:-Move sheets are generated, independently

verified, and approved by qualified individuals

-Blocking devices or trash cans are placed in locations that are face-adjacent tolocations approved for the storage of fresh fuel-The fuel movers will verify that fresh fuel is not placed in face-adjacent locations prior to completing each fuel moveSufficient rigor is placed into the generation of move sheets, execution of fuel movement, andmaintenance of SEP maps that the likelihood of an assembly being misplaced in the SEP issmall. The misplacement of multiple fuel assemblies is less probable.

Therefore, the approachused in the analysis is appropriately conservative.

The impact of placing multiple fresh fuelassemblies in face-adjacent locations is not evaluated in the analysis because this event is notconsidered credible.

A regulatory commitment regarding the implementation of procedural controls to requireverification that fresh fuel assemblies are not placed face-adjacent to one another beforecompleting a fuel move is provided in Attachment 4.3.2 New Fuel Storage and Fuel Transfer Equipment AnalysisA criticality safety analysis was performed to support operation of the NFS racks, the IFSR, thenew fuel elevator, and the fuel upender and transfer machine.

When discussing the new fuelelevator, and the fuel upender and transfer machine together, they are referred to as the fuelhandling equipment (FHE). The existing NFS racks, IFSR, and FHE were evaluated to confirmthat each system maintains subcriticality while performing its designed purposes.

3.2.1 Storage and Equipment Description N.ew Fuel Storagqe DesigqnThe NFS rack assemblies are made up of individual racks similar to those shown in UFSARFigure 9.1-1. A minimum edge-to-edge spacing between fuel assemblies is maintained inadjacent rows. This spacing is the minimum value after allowances are made for rack fabrication tolerances and the predicted deflections resulting from postulated accident conditions.

The stainless steel construction of the storage racks is compatible with the water and thezirconium-clad fuel. The top structure of the racks is designed such that there is no openingbetween adjacent fuel cavities that is as large as the cross-section of the fuel bundle. Inaddition, the outer structure of the racks precludes the inadvertent placement of a bundleagainst the rack closer than the prescribed edge-to-edge spacing.17 Enclosure Description and Assessment of Proposed License Amendment Two concrete storage cavities are utilized for NFS. Each cavity is approximately 8 feet by23 feet and contains 45 fuel assemblies in stainless steel racks. Three racks are installed ineach cavity, forming a 3x15 array of fuel assemblies.

The rack structure provides at least 10 inches between the top of the active fuel and the top ofthe rack to preclude criticality in case a fuel assembly is dropped into a horizontal position onthe top of the rack. The NFS racks and facilities are qualified as Seismic Category I and willsurvive a safe shutdown earthquake without loss of safety function.

The following postulated accidents were considered in the design of the NFS racks:* Flooding

-complete immersion of the entire storage array in pure, unborated, roomtemperature water* Envelopment of the entire array in a uniform density aqueous foam or mist of optimumdensity that maximizes the reactivity of the finite array (a condition that could result fromfirefighting)

• A fuel assembly dropped from a height of 4.5 ft onto the rack that falls horizontally across the top of the rack* Tensile load of 5000 lbs on the rackAlthough the above accident conditions have been postulated, the FHE, NFS racks, and thebuilding arrangement are designed to minimize the possibility of these accidents and the effectsresulting from these accidents.

Intermediate Fuel Storagqe RackThe IFSR is a four-cavity fuel storage rack in a lx4 array designed as an intermediate storagelocation for fuel bundles during refueling.

The rack is located in the containment adjacent to thecore support barrel laydown area, which provides access to the refueling machine for insertion and removal of fuel bundles.Each cavity in the IFSR is a stainless steel can 8.69 inches on a side. The cavities areseparated by a fuel center-to-center pitch of 18.56 inches. Each of the cavities is open at thebottom to provide thermal cooling for the worst case fuel bundle. The rack structure is designedto maintain keff < 0.95 by assuring under all normal and accident conditions, which includesSSE, that the minimum edge distance is not violated and that a fuel bundle cannot violate the12-inch minimum stand-off distance around the cavities.

New fuel may be stored in the IFSR before being moved into the core. Partially spent fuel maybe moved out of the core and stored temporarily in the IFSR to provide spaces for fuel shuffling.

Spent fuel may be stored in the IFSR before being sent to the SFP.Fuel Upender and Transfer MachineThe transfer

machine, or carriage, conveys the fuel assemblies through the transfer tube. Twofuel assembly cavities are provided in the fuel carriage to reduce overall fuel handling time. Afterthe refueling machine deposits a spent fuel bundle in the open cavity, it only has to moveapproximately one foot to pick up the new fuel assembly, which was brought from the fuelbuilding in the other cavity. The handling operation in the fuel building is similar.

The dual cavity18 Enclosure Description and Assessment of Proposed License Amendment arrangement permits both fuel handling machines to travel fully loaded at all times. Fuelassemblies are placed on the transfer carriage in a vertical

position, lowered to the horizontal

position, moved through the fuel transfer tube on the transfer

carriage, and then restored to thevertical position.

Wheels support the carriage and allow it to roll on tracks within the transfer*tube. The track sections at both ends of the transfer tube are mounted on the upendingmachines to permit the carriage to be properly positioned at the limits of its travel.An upending machine is provided at each end of the transfer tube. Each machine consists of aStructural support base from which is pivoted an upending straddle frame that engages the two-cavity fuel carrier.

Hydraulic cylinders attached to the upending frame that engages the supportbase rotate the fuel carrier between the vertical and horizontal position.

A third fuel assemblywas modeled five inches from the transfer carriage to allow for the presence of an additional fuelassembly no closer than five inches from the carriage.

(Appendix B of Attachment 8)New Fuel ElevatorThe new fuel elevator is utilized to lower new fuel from the operating floor to the bottom of thepool where it is grappled by the spent fuel handling tool. The elevator is powered by a cablewinch and fuel is contained in a simple support structure whose wheels are captured in tworails.3.2.2 New Fuel Criticality Safety AnalysisAcceptance CriteriaThe objective of the criticality safety analysis is to ensure that the fuel storage operations arewithin the bounds 10 CFR 50.68(b)(2) and 50.68(b)(3):

• The estimated ratio of neutron production to neutron absorption and leakage (keff) of thefresh fuel in the storage racks shall be calculated assuming the racks are loaded withfuel of the maximum fuel assembly reactivity and flooded with unborated water and mustnot exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent suchflooding or if fresh fuel storage racks are not used.* If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racksare assumed to be loaded with fuel of the maximum fuel assembly reactivity and filledwith low-density hydrogenous fluid, the keff corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. Thisevaluation need not be performed if administrative controls and/or design featuresprevent such moderation or if fresh fuel storage racks are not used.Design ApproachCompliance is shown for the NFS racks, IFSR, and EHE by demonstrating that the system does not exceed 0.95 at a 95 percent probability with a 95 percent confidence level. Aconservative combination of best estimate and bounding values has been selected to model thefuel in the analysis to ensure that fuel represented by the proposed TS is less reactive than thefuel modeled in the analysis.

19 Enclosure Description and Assessment of Proposed License Amendment Computer CodesThe analysis methodology employs SCALE 6.1.2 (Reference 6.3) with the 238-group cross-section library based on ENDF/B-VII.

All analyses performed used the "Fresh fuel withoutAbsorber' validation suite. KENO-Va is used to determine the absolute reactivity of fresh fuelassemblies in the NFS.The validation of the ENDF/B-VII library with the SCALE 6.1.2 CSAS5 module is documented inAppendix A of Attachment

8. The code validation shows that SCALE 6.1.2 is an accurate toolfor calculation of keff for the applications in this LAR. The benchmark calculations utilize thesame computer plafform and cross-section libraries that are used for the design basiscalculations.

Limitingq Fuel Desigqn Selection There are four potentially limiting fuel designs that have been used at PVNGS. The VAP designis currently in use on site and NGF is planned for use in the future. Therefore, both VAP andNGF designs are considered the two potentially limiting fuel designs.

The STD and ANP fueldesigns do not need to be addressed for these calculations because they are bounded by theVAP fuel design. Additional information is provided in Section B.4.2 of Attachment 8.Treatment of ConcreteConcrete is a material which has a large variety of different potential compositions, all of whichcan be labeled as "concrete."

Attachment 8 references the most limiting design for eachsituation.

The analysis has used a bounding treatment for concrete and the methodology remains conservative throughout the life of the concrete.

Additional information is provided inSection B.4.3 of Attachment 8.Biases and Uncertainties Reactivity biases are known variations between the real and analyzed system and theirreactivity impact is added directly to the calculated keff. Uncertainties are random dispersions around a nominal, measured quantity.

Their impact is added to the calculated keff as the squareroot of the sum of the squares of the uncertainties.

The following biases and uncertainties areaccounted for in the analysis.

A detailed discussion of biases and uncertainties is provided inSection B.4.5 of Attachment 8.* Reactivity effect of manufacturing tolerances

• Structural material presence* Eccentric fuel assembly positioning

• Uncertainty in the predictive capability of SCALE 6.1.2 and the associated cross-section library* Temperature bias for operating temperature range* Planar enrichment bias20 Enclosure Description and Assessment of Proposed License Amendment New Fuel Storagqe Rack Criticality Safety AnalysisThe criticality safety analysis for the NFS rack consists of determining the limiting fuel designunder both the fully flooded and optimum moderation condition.

Biases and uncertainties forboth fully flooded and optimum moderation conditions are calculated using the limiting fueldesign. The best estimate keff of the NES rack under both full density water and optimummoderation conditions is less than the target keff. This demonstrates that the NFS rack complieswith the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.6 ofAttachment 8.The analysis of the NES rack has demonstrated that it can be operated in its design capacitywithout risk of exceeding the maximum reactivity imposed by regulation.

The analysis supportsuse of these components up to a maximum radially averaged enrichment of 4.65 weight percentU-235. All fuel used to date at PVNGS has an initial enrichment of < 4.55 weight percent U-235.Intermediate Fuel Storaqe Rack Criticality Safety AnalysisThe criticality safety analysis for the IFSR consists of determining the target keff for the IFSR,then confirming that the best estimate system keff (plus 2 o) is below the target keff with thelimiting fuel design. The analysis uses the NGF design because the NGF is more reactive withfull density water than VAP fuel. The biases and uncertainties are also calculated using the NGFdesign. The best estimate keff of the IFSR is less than the target keff, which demonstrates thatthe IFSR complies with the requirements of 10 CFR 50.68. Additional information is provided inSection B.4.7 of Attachment 8.Fuel Upender.

Transfer

Machine, and New Fuel Elevator Criticality Safety AnalysisThe criticality safety analysis for the fuel upender and transfer

machine, and the analysis for thenew fuel elevator, demonstrate that they can be used with fresh 4.65 weight percent U-235 fuelwithout exceeding a keff of 0.95 at a 95 percent probability, 95 percent confidence level. Thedesign basis fuel is the NGF design. The best estimate keff of the fuel upender and transfermachine, and the new fuel elevator, is less than the target which demonstrates compliance with the requirements of 10 CFR 50.68. Additional information is provided in Section B.4.8 ofAttachment 8.3.3 Spent Fuel Pool Transition PlanThe SEP transition will be conducted over a total lapsed time of approximately 24 months with aschedule based on the unit refueling outages.

Therefore, APS will insert two sets of TS and TSBases pages during implementation.

One set will be labeled "Before SEP transition" and theother set will be labeled "After SFP transition."

The Spent Fuel Pool Transition Plan is based on a SEP module-by-module transition scheme.Each SEP module that has not been transitioned is governed by the 'Before SEP transition" pages. As each module in a SEP is transitioned to the new configuration, the Shift Manager willmake an entry in the control room log and declare that module as "transitioned."

That particular module is then governed by the "After SEP transition" pages. When all three units have beentransitioned, APS will submit an administrative TS change to remove the "before" and "after"pages, and insert the final pages.21 Enclosure Description and Assessment of Proposed License Amendment APS will transition to the proposed TS 3.7.17 in each of the three units in the following manner:1. Move fuel assemblies as needed in order to neutronically decouple one module from thebalance of the SEP. Analysis demonstrates that one row of empty cells is enough todecouple modules in the SEP.2. Perform a shuffle of the decoupled module, including installation of NETCO-SNAP-IN rack inserts.a. Some modules may need to be completely emptied of fuel assemblies.

Theassemblies that must be moved may be stored in the appropriate regions of therest of the SEP.b. Fuel assemblies that already meet the new TS 3.17.17 requirements may notneed to be shuffled.

3. Upon completion of the fuel shuffle and NETCO-SNAP-IN rack insert installation, theShift Manager will declare the module has transitioned to the new TS 3.7.17 and enterthis information into the control room log.4. Perform additional fuel shuffles, as needed, in order to move fuel from other parts of theSEP to the recently transitioned module.5. Repeat steps 1 through 4 for all 17 modules.6. Perform a 100 percent pool verification to confirm the following

a. Fuel has been properly moved as confirmed by a 100 percent serial numbercheckb. Stainless steel L-Jnserts are in the locations assumed in the analysis of recordc. NETCO-SNAP-IN rack inserts are in their assumed locations

d. Blocking devices are in their assumed locations A generic Westinghouse study (Reference 6.6) investigated the adequacy of assuming adistance of one cell pitch for neutronic isolation.

The study included NGF, which was determined to be the limiting fuel type used in the PVNGS SEP criticality safety analysis.

The results of thestudy concluded that a distance of approximately 10 cm (3.94 inches) of water was adequate forneutronic decoupling.

Given the required separation distance for neutronic decoupling, fuelassemblies separated by a single cell pitch or more at PVNGS are neutronically decoupled.

Once the Spent Fuel Pool Transition Plan has been started in a particular unit, the insertinstallation and SFP transition shall be executed in a deliberate, safe, and controlled manneruntil complete in that unit. The transition to the new SEP configuration will be completed in allthree units in accordance with the Spent Fuel Pool Transition Plan within two years of the NRCapproval date of the amendment or by December 31, 2019, whichever is later. A regulatory commitment regarding transition plan implementation is provided in Attachment 4.4.0 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements The regulations in 10 CFR 50.36(c)(2)(ii)(B),

Limiting conditions for operation, state:22 Enclosure Description and Assessment of Proposed License Amendment Criterion

2. A process variable, design feature, or operating restriction that is aninitial condition of a design basis accident or transient analysis that eitherassumes the failure of or presents a challenge to the integrity of a fission productbarrier.Technical Specification (TS) 3. 7.17 currently meets this requirement and will continue to meetthis requirement after the proposed changes are approved and implemented.

The regulations in 10 CFR 50.36(c)(4),

Design features, state:Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, wouldhave a significant effect on safety and are not covered in categories described inparagraphs (c)(1), (2), and (3) of this section.TS 4.3.1 currently meets this requirement and will continue to meet this requirement after theproposed changes are approved and implemented.

The regulations in 10 CFR 50.68, Criticality accident requirements, specifically 10 CFR50.68(b)(1) state:Plant procedures shall prohibit the handling and storage at any one time of morefuel assemblies than have been determined to be safely subcritical under themost adverse moderation conditions feasible by unborated water.This requirement is currently met by existing PVNGS fuel handling procedures and will continueto be met by the same procedures after the proposed changes are approved and implemented.

The regulations in 10 CER 50.68(b)(2) state:The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racksare loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or designfeatures prevent such flooding or if fresh fuel storage racks are not used.The regulations in 10 CFR 50.68(b)(3) state:If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racksare assumed to be loaded with fuel of the maximum fuel assembly reactivity and filledwith low-density hydrogenous fluid, the k-effective corresponding to this optimummoderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or designfeatures prevent such moderation or if fresh fuel storage racks are not used.The regulations in 10 CFR 50.68(b)(4) state:If no credit for soluble boron is taken, the k-effective of the spent fuel storageracks loaded with fuel of the maximum fuel assembly reactivity must not exceed0.95, at a 95 percent probability, 95 percent confidence level, if flooded withunborated water. If credit is taken for soluble boron, the k-effective of the spent23 Enclosure Description and Assessment of Proposed License Amendment fuel storage racks loaded with fuel of the maximum fuel assembly reactivity mustnot exceed 0.95, at a 95 percent probability, 95 percent confidence level, ifflooded with borated water, and the k-effective must remain below 1.0(subcritical),

at a 95 percent probability, 95 percent confidence level, if floodedwith unborated water.The requirements in 10 CFR 50.68(b) cited above are met by the nuclear criticality safetyanalyses provided in WCAP- 18030-P.

The results of the criticality analysis form the basis of theproposed TS 3.7.17 changes.

TS 3.7.17 currently meets these requirements and will continue tomeet these requirements after the proposed changes are approved and implemented.

The regulations in 10 CFR 50.68(b)(7) state:The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five(5.0) percent by weight.TS 4.3.1.2 currently meets this requirement and will continue to meet this requirement after theproposed changes are approved and implemented.

The regulations in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear PowerPlants, Criterion 62, Prevention of criticality in fuel storage and handling, state:Criticality in the fuel storage and handling system shall be prevented by physicalsystems or processes, preferably by use of geometrically safe configurations.

TS 3. 7.17 currently meets this requirement and will continue to meet this requirement after theproposed changes are approved and implemented.

The guidance in DSS-ISG-2010-01 is to be used by NRC staff to review nuclear criticality safetyanalyses for the storage of new and spent nuclear fuel as they apply to applications for licenseamendments submitted after September 29, 2011.This license amendment request and WCAP-18030-P were developed using the guidance ofDSS-ISG-2010-01.

4.2 Precedent

The analysis methodology for the site-specific criticality analysis employs the PARAGON code,which is approved for use by the NRC (Reference 6.4).4.:3 Significant Hazards Consideration As required by 10 CFR 50.91(a),

Notice for Public Comment, an analysis of the issue of nosignificant hazards consideration using the standards in 10 CFR 50.92, Issuance ofAmendment, is presented below:1. Does the proposed amendment involve a significant increase in the probability orconsequences of an accident previously evaluated?

Response:

No.24 Enclosure Description and Assessment of Proposed License Amendment The proposed amendment would modify the Palo Verde Nuclear Generating Station(PVNGS) Technical Specifications (TS) to incorporate the results of an updated criticality safety analysis for both new fuel and spent fuel storage.

The revised criticality safetyanalysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previouscriticality safety analysis.

The proposed amendment does not change or modify the fuel, fuel handling processes, number of fuel assemblies that may be stored in the spent fuel pool (SFP), decay heatgeneration rate, or the SFP cooling and cleanup system. The proposed amendment wasevaluated for impact on the following previously evaluated events and accidents:

• fuel handling accident (FHA)* fuel misload event* SEP boron dilution event* seismic event* loss of SEP cooling eventImplementation of the proposed amendment will be accomplished in accordance with theSpent Fuel Pool Transition Plan and does not involve new fuel handling equipment orprocesses.

The radiological source term of the fuel assemblies is not affected by theproposed amendment request.

The EHA radiological dose consequences associated withfuel enrichment at this level are addressed in the PVNGS Updated Final Safety AnalysisReport (UFSAR) Section 15.7.4 and remain unchanged.

Therefore, the proposedamendments do not significantly increase the probability or consequences of a FHA.Operation in accordance with the proposed amendment will not change the probability of afuel misload event because fuel movement will continue to be controlled by approved fuelhandling procedures.

Although there will be additional allowable storage arrays defined bythe amendment, the fuel handling procedures will continue to require identification of theinitial and target locations for each fuel assembly that is moved. The consequences of a fuelmisload event are not changed because the reactivity analysis demonstrates that the samesubcriticality criteria and requirements continue to be met for the limiting fuel misload event.Operation in accordance with the proposed amendment will not change the probability orconsequences of a boron dilution event because the systems and events that could affectSFP soluble boron concentration are unchanged.

The current boron dilution analysisdemonstrates that the limiting boron dilution event will reduce the boron concentration fromthe TS limit of 2150 ppm to 1900 ppm. This leaves sufficient margin to the 1460 ppmcredited by the SFP criticality safety analysis.

The analysis confirms that the time needed fordilution to reduce the soluble boron concentration is greater than the time needed for actionsto be taken to prevent further dilution.

Operation in accordance with the proposed amendment will not change the probability of aseismic event since there are no elements of the updated criticality analysis that influence the occurrence of a seismic event. The consequences of a seismic event are notsignificantly increased because the forcing functions for seismic excitation are not increased and because the mass of storage racks with NETCO-SNAP-IN inserts is not appreciably 25 Enclosure Description and Assessment of Proposed License Amendment increased.

Seismic analyses demonstrate adequate stress levels in the storage racks wheninserts are installed.

Operation in accordance with the proposed amendment will not change the probability of aloss of SEP cooling event because the systems and events that could affect SEP cooling areunchanged.

The consequences are not significantly increased because there are nochanges in the SFP heat load or SEP cooling systems, structures, or components.

Furthermore, conservative analyses indicate that the current design requirements andcriteria continue to be met with the NETCO-S NAP-IN inserts installed.

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

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

Response:

No.The proposed amendment would modify the PVNGS TS to incorporate the results of anupdated criticality safety analysis for both new fuel and spent fuel storage.

The revisedcriticality safety analysis provides an updated methodology that allows credit for neutronabsorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in the previous criticality safety analysis.

The proposed amendment does not change or modify the fuel, fuel handling processes, number of fuel assemblies that may be stored in the pool, decay heat generation rate, or theSEP cooling and cleanup system. The effects of operating with the proposed amendment are listed below. The proposed amendment was evaluated for the potential of each effect tocreate the possibility of a new or different kind of accident:

• addition of inserts to the SEP storage racks* additional weight from the inserts* new storage patterns* displacement of SEP water by the inserts,Each NETCO-SNAP-IN insert will be placed between a fuel assembly and the storage cellwall, taking up some of the space available on two sides of the fuel assembly.

Analysesdemonstrate that the presence of the inserts does not adversely affect spent fuel cooling,seismic capability, or subcriticality.

The aluminum and boron carbide materials ofconstruction have been shown to be compatible with nuclear fuel, storage racks, and SEPenvironments, and generate no adverse material interactions.

Therefore, placing the insertsinto the SEP storage racks cannot cause a new or different kind of accident.

Operation with the added weight of the NETCO-SNAP-IN inserts will not create a new ordifferent accident.

The analyses of the racks with NETCO-SNAP-IN inserts installed demonstrate that the stress levels in the rack modules continue to be considerably less thanallowable stress limits. Therefore, the added weight from the inserts cannot cause a new ordifferent kind of accident.

26 Enclosure Description and Assessment of Proposed License Amendment Operation with the proposed fuel storage patterns will not create a new or different kind ofaccident because fuel movement will continue to be controlled by approved fuel handlingprocedures.

These procedures continue to require identification of the initial and targetlocations for each fuel assembly that is moved. There are no changes in the criteria ordesign requirements pertaining to fuel storage safety, including subcriticality requirements.

Analyses demonstrate that the proposed storage patterns meet these requirements andcriteria with adequate margins.

Therefore, the proposed storage patterns cannot cause anew or different kind of accident.

Operation with insert movement above stored fuel will not create a new or different kind ofaccident.

The insert with its handling tool weighs less than the weight of a single fuelassembly.

Single fuel assemblies are routinely moved safely over fuel assemblies and thesame level of safety in design and operation will be maintained when moving the inserts.The installed rack inserts will displace a negligible quantity of the SEP water volume andtherefore will not reduce operator response time to previously-evaluated SFP accidents.

The accidents and events previously analyzed remain bounding.

Therefore, the proposedamendment does not create the possibility of a new or different kind of accident from anyaccident previously evaluated.

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

No.The proposed amendment would modify the TS to incorporate the results of an updatedcriticality safety analysis for both new fuel and spent fuel storage.

The revised criticality safety analysis provides an updated methodology that allows credit for neutron absorbing NETCO-SNAP-IN rack inserts and corrects non-conservative input assumptions in theprevious criticality safety analysis.

It was evaluated for its effect on current margins of safetyas they relate to criticality, structural integrity, and spent fuel heat removal capability.

The margin of safety for subcriticality required by 10 CFR 50.68(b)(4) is unchanged.

Newcriticality analyses confirm that operation in accordance with the proposed amendment continues to meet the required subcriticality margins.The structural evaluations for the racks and spent fuel pool with NETCO-SNAP-IN insertsinstalled show that the rack and SEP are unimpaired by loading combinations during seismicmotion, and there is no adverse seismic-induced interaction between the rack and NETCO-SNAP-IN inserts.The proposed amendment does not affect spent fuel heat generation, heat removal from thefuel assembly, or the SEP cooling systems.

The effects of the NETCO-SNAP-IN inserts arenegligible with regards to volume of water in the pool, flow in the SEP rack cells, and heatremoval system performance.

The addition of a Spent Fuel Pool Rack Neutron Absorber Monitoring program (proposed TS5.5.21) provides a method to identify potential degradation in the neutron absorber materialprior to challenging the assumptions of the criticality safety analysis related to the material.

Therefore, the addition of this monitoring program does not reduce the margin of safety;27 Enclosure Description and Assessment of Proposed License Amendment rather it ensures th'e margin of safety is maintained for the planned life of the spent fuelstorage racks.Therefore, the proposed amendment does not involve a significant reduction in the marginof safety.4.4 Conclusion APS concludes that operation of the facility in accordance with the proposed amendment doesnot 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 theconsiderations discussed above, (1) there is reasonable assurance that the health and safety ofthe public will not be endangered by operation in the proposed manner, (2) such activities willbe conducted in compliance with the Commission's regulations, and (3) the issuance of theamendment will not be inimical to the common defense and security or the health and safety ofthe public.5.0 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement withrespect to installation or use of a facility component located within the restricted area, as definedin 10 CFR 20, Standards for Protection Against Radiation.

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

offsite, or (iii) asignificant increase in individual or cumulative occupational radiation exposure.

Accordingly, theproposed amendment meets the eligibility criterion for categorical exclusion set forth in10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51 .22(b), no environmental impactstatement or environmental assessment need be prepared in connection with the proposedamendment.

6.0 REFERENCES

6.1 Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3(Proprietary),

WCAP-1 8030-P, Revision 0, September 2015.6.2 Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools,DSS-ISG-201 0-01, Revision 0, Nuclear Regulatory Commission Division of SafetySystems, Rockville, MD, September 29, 2011. (ML1 10620086) 6.3 Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis andDesign, ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge,TN, June 2011.6.4 M. Ouisloumen, H. Huria, et al, Qualification of the Two-Dimensional Transport CodePARAGON, WCAP-16045-P-A, Revision 0, Westinghouse Electric Company LLC,Monroeville, PA, August 2004.28 Enclosure Description and Assessment of Proposed License Amendment 6.5 C. V. Parks, et al, Review and Prioritization of Technical Issues Related to BumupCredit for LWR Fuel, NUREG/CR-6665, Oak Ridge National Laboratory, Oak Ridge,TN, February 2000.6,6 Letter, J. Gresham (WEC) to NRC, Responses to Requests for Additional Information from the Review of WCAP- 1 7483-PA/WCAP-1 7483-NP, Revision 0, 'Westinghouse Methodology for Spent Fuel Pool and New Fuel Rack Criticality Safety Analysis,'

LTR-NRC-15-60, dated July 20, 2015.29 Enclosure

,Description and Assessment of Proposed License Amendment ATTACHMENT 1Marked-up Technical Specifications Pages(Pages Provided for Before and After SEP Transition) 3.7.17 3.7.17-23.7.17-33.7.17-44.0-24.0-35.5-19 SBefore SFP transitionI Spent Fuel Assembly Storage3.7.173.7 PLANT SYSTEMS3.7.17 Spent Fuel Assembly StorageLCO 3.7.17APPLICABILITY:

The combination of initial enrichment, burnup, and decaytime of each fuel assembly stored in each of the fourregions of the fuel storage pool shall be within theacceptable burnup domain for each region as shown in Figures3.7.17-1, 3.7.17-2, or 3.7.17-3, and described inSpecification 4.3.1.1.Whenever any fuel assembly is stored in the fuel storagepool.ACTIONS__________________________

CONDITION REQUIRED ACTION COMPLETION TIMEA. Requirements of the A.1------NOTE----

LCO not met. LCD 3.0.3 is notapplicable.

Initiate action to Immediately move the noncomplying fuel assembly into anappropri ate region.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior toenrichment, burnup, and decay time of the storing thefuel assembly is in accordance with Figures fuel assembly3.7.17-1, 3.7.17-2, or 3.7.17-3, and in the fuelSpecification 4.3.1.1.

storage pool.PALO VERDE UNITS 1,2,33.7.17-IPAL VEDE NIT 1,,3 .7.7-1AMENDMENT NO. 117, 1£ IBefore SFP transition]I Spent Fuel Assembly Storage3.7.17Figure 3.7.17-1ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT forRegion 2060SSSS0ACC EPTABL["

for Reg on 215000500I-NOT A CEPTA LE for R gion 2SSSSSSSSSS0SSSSSSS_________________

_________________

9SSSSSSSSSNote: This curve assumes ero decay time.____________

~1~S(11.52.02.53.0 3.5Initial Enrichment, weight %4.04.5

• 5.04.80%limitingenrichment PALO VERDE UNITS 1,2,33.7.17-2AMENDMENT NO. 117, !2

[Before SFP transition Spent Fuel AssemblyStorage3.7.17Figure 3.7.17-2ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT forRegion 3(at decay times from 0 to 20 years)4500040000350003000025000>20000t'0Eci0ACCEP' ABLE for Region 3 "o,.S//NOT ACq;EPTABI.

E for RegionS0ThOUU+J +

IS0SS(S-a---SS5000~~.2Noe Asnnent and current diS0Scay lime.:ly e/igible for Regi jm 3 if actualIBU

> 3U requirement for given initial endchFi, , , , .....m1.5 2.0 2.5 3.0 3.5 4.0 4.5

• 5.0Initial Enrichment, weight % 4.80%limitingDecaylime[

-U-1-5years

--4--2Oyears enrichment PALO VERDE UNITS 1,2,33..73AMNETNO 3.7.17-3AMENDMENT NO.

I Before SFP transition Spent Fuel Assembly Storage3.7.17Figure 3.7.17-3ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT

~forRegion 4(at decay times from C to 20 years)I-0E~0Ea)0)3.0 3.5Initial Enrichment, weight %limitingDecaylimeI

--0years 11-years

--&.-lOyears

-~-'-15 years --o-20years enrdchment PALO VERDE UNITS 1,2,3 371- MNMN O3.7.17-4AMENDMENT NO.

SAfter SFP transition Spent Fuel Assembly Storage3.7.173.7 PLANT SYSTEMS 13.7.17-1 through 3.7.17-5.I 3.7.17 Spent Fuel Assembly StorageLCO 3.7.11 The combination of initialnichment, burnup, and decaytime of each fuel assemblytreine=hftef'-

acceptable burnup domain for each region as shown in Figures') 7 17 1 ') 7 17 ') ,-~-i Q 7 17 '2 -~A A-,-.~k,-A

-vsSpecification 1.3.1.1.APPLICABILITY:

Whenever any fuelpool.assembly is stored in the fuel storageACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Requirements of the A.1------NOTE--

--LCO not met. LCO 3.0.3 is notapplicable.

Initiate action to Immediately move the noncomplying fuel assembly into anappropriate region.SURVEILLANCE REQUIREMENTS_________

SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior toenrichment, burnup, and decay time of the storing thefuel assembly is in accordance with fuel assembly...and in the fuelSpecification 4mi.3.1.1

... storage pool.ITables 3.7.17-1 through 3.7.17-5, Figure 3.7.17-PALO VERDE UNITS 1,2,33.7.17-1PAL VEDE NIT 1,,3 .7.7-1AMENDMENT NO. I147, 2 Ilnsert new Tables 3.7.17-1 through 3.7.17-5 andjFigure 3.7.17-1 (total 6 pages) here.IIAfter SFP transitionI Spent Fuel Assembly Storage3.7.17FiguJren 3.7.17 1Acc(rkArnl

\J l \lII-1CI IC TkITTT Al FI-dITriUMCIdT PALO VERDE UNITS 1,2,397179PAL VRD UIT 12, 37.72AMENDMENT NO. 11.7, 2 SAfter SFP transition Spent Fuel Assembly Storage3.7.17limiting--me '-O years --U-5 years -* 10 years -U-I--l5 years --4-20 years enrichment PALO VERDE UNITS 1,2,33. 13AMNETNO 1597]79AMENDMENT NO.

SAfter SFP transitionI Spent Fuel Assembly Storage3.7.17Figure,3.7.17 3-ACSE'MILY VU NUP VERSUS TIT-I-AI ENRT/ICHMEITT

'4,~uuuU45000SS40000 __ __ _rSACCEIRegion 4350002500000..~2O0001500010000I 0SS_________________

_________

ISSSS__________________

___________________

S0SSNOT AC E ABLE or Region ____ __SSSSS_________________

_________________

_________________

_________

S______/SSSS0/ ______ + 4-~---f5000//N______ 4- + 4-SS0S000caYthne.

0mq~kemwl fortyB~efor Reg M 4&fachiBU

>-hiaI enddi rdw caidwmt1.2.0 2.5 3.0 3.5 4.0 4.5

• 5Initial Enrichment, weight % 4.80%-ieI-4-0 years --11-5 years 15 years --4-20Oyears lenicmetnt PALO VERDE UNITS 1,2,33..7AMN ETN. §Q717AAMENDMENT NO.

[After SFP transition]

Spent Fuel Assembly Storage3.7.17Table 3.7.17-1Fuel RegionsRanked by Reactivity Fuel Region 1 Highest Reactivity (See Note 2)Fuel Region 2Fuel Region 3Fuel Region 4Fuel Region 5Fuel Region 6 Lowest Reactivity Notes:1. Fuel Regions are defined by assembly average burnup, initial enrichment' and decay time asprovided by Table 3.7.17-2 through Table 3.7.17-5.

2. Fuel Regions are ranked in order of decreasing reactivity, e.g., Fuel Region 2 is less reactivethan Fuel Region 1, etc.3. Fuel Region 1 contains fuel with an initial maximum radially averaged enrichment up to4.65 wt% 235U. No burnup is required.

4. Fuel Region 2 contains fuel with an initial maximum radially averaged enrichment up to4.65 wt% 235U with at least 16.0 GWd/MTU of bumup.5. Fuel Regions 3 through 6 are determined from the minimum burnup (BU) equation andcoefficients provided in Tables 3.7.17-2 through 3.7.17-5.

6. Assembly storage is controlled through the storage arrays defined in Figure 3.7.17-1.

7. Each storage cell in an array can only be populated with assemblies of the Fuel Region definedin the array definition or a lower reactivity Fuel Region.SInitial Enrichment is the nominal 235U enrichment of the central zone region of fuel, excluding axial blankets, priorto reduction in 235U content due to fuel depletion.

If the fuel assembly contains axial regions of different 235Ujenrichment values, such as axial blankets, the maximum initial enrichment value is to be utilized.

[After SFP transition[

Spent Fuel Assembly Storage3.7.17Table 3.7.17-2Fuel Region 3: Burnup Requirement Coefficients Coefficients DecayTime (yr.) A1 A2 A3 A40 -1.5473 15.5395 -39.0197 24.11215 -1.4149 13.9760 -33.6287 18.336910 -1.3012 12.6854 -29.2539 13.687915 -1.0850 10.4694 -22.1380 6.367320 -0.9568 9.1487 -17.9045 2.0337Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meetthe requirements of a Fuel Region, the assembly burnup must exceed the "minimum burnup"(GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated from the following equation:

BU =Al

• En3 + A2

• En2 + A3

• En + A42. Initial enrichment, En, is the maximum radial average 235U enrichment.

Any En value between2.55 wt% 235U and 4.65 wt% 235U may be used. Burnup credit is not required for an En below2.55 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decaytime of greater than 20 years.

[After SFP transition Spent Fuel Assembly Storage3.7.17Table 3.7.17-3Fuel Region 4: Burnup Requirement Coefficients Coefficients DecayTime (yr.) A1 A2 A3 A40 0.4260 -6.2766 40.9264 -54.68135 0.2333 -4.1!545 32.9080 -46.116110 0.4257 -6.2064 39.0371 -51.588915 0.53 15 -7.3777 42.5706 -54.752420 0.5222 -7.3897 42.6587 -54.8201Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meetthe requirements of a Fuel Region, the assembly bumup must exceed the "minimum burnup"(GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated from the following equation:

BU=AI

• En3 +A2 *En2+-bA3 *En +A42. Initial enrichment, En, is the maximum radial average 235U enrichment.

Any En value between1.75 wt% 235U and 4.65 wt% 235U may be used. Bumnup credit is not required for an En below1.75 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decaytime of greater than 20 years.

[After SEP transition Spent Fuel Assembly Storage3.7.17Table 3.7.17-4Fuel Region 5: Burnup Requirement Coefficients Decay Coefficients Time(yr.) A1 A2 A3 A40 -0.1114 -0.4230 20.9136 -32.85515 -0.1232 -0.4463 20.8337 -32.606810 -0.2357 0.4892 18.0192 -30.004215 -0.1402 -0.4523 20.3745 -31.756520 -0.0999 -0.8152 21.0059 -31.9911Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly tomeet the requirements of a Fuel Region, the assembly burnup must exceed the "minimum bumup"(GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated from the following equation:

BU =A1

• En3 + A2

• En2 + A3

• En + A42. Initial enrichment, En, is the maximum radial average 235U enrichment.

Any En value between1.65 wt% 235U and 4.65 wt% 235U may be used. Burnup credit is not required for an En below1.65 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decaytime of greater than 20 years.

[After SFP transition Spent Fuel Assembly Storage3.7.17Table 3.7.17-5Fuel Region 6: Burnup Requirement Coefficients Decay Coefficients Time(yr.) A1 A2 A3 A40 0.7732 -9.3583 49.6577 -54.68475 0.7117 -8.4920 45.1124 -49.728210 0.6002 -7.2638 40.2603 -44.934815 0.5027 -6.2842 36.6715 -41.493420 0.2483 -3.7639 28.8269 -34.6419Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

For instance, no additional allowance for bumnup uncertainty or enrichment uncertainty is required.

For a fuel assembly tomeet the requirements of a Fuel Region, the assembly bumup must exceed the "minimum burnup"(GWd/MTU) given by the curve fit for the assembly "decay time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated from the following equation:

BU =A1

• En3 + A2

• En2 + A3

• En + A42. Initial enrichment, En, is the maximum radial average 235U enrichment.

Any En value between1.45 wt% Z3U and 4.65 wt% 23U may be used. Burnup credit is not required for an En below1.45 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based on decay time.4. The 20-year coefficients must be used to calculate the minimum BU for an assembly with a decaytime of greater than 20 years.

SAfter SFP trransition Spent Fuel Assembly Storage3.7.17Figure 3.7.17-1Allowable Storage ArraysFwou Region 6 assemblies (6) Tw steckrbagded cells cotain abtlocess stells L(ise). TheRgo1. Thsembishaded loathionsa indicatehelhcoti a stainless steel L-insert.NoETOSA-N 2.o Ae block1asedbis()cekrbaddwt w cells(X contains lcing dvcanolyw terainshe actiefulrein 3TC. NTheRgo-n1assemlNinets must bc oientedl winthe saedrcinaah stainless steel L-inserts.Eer 4 NTC -N PN isrsaeolloaeincells without a stainless steel L-inetms oti EC -N P1insert.

5-nsr. Anyhel egontann3 afe assemblyyr iCsa iseaiennemt (aerflld cell intinn aNTCall-Niset stoageayrys 6.e AnR traearagoaion deintdfrafe assembly may cbeckreplaced with noren-gon4fssssiele).Th mein2atserial.

n h ignlylctdRein4asml r ahi I Before SFP transitionI Design Features4.04.0 DESIGN FEATURES (continued) 4.3 Fuel Storage4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall bemaintained with:a. Fuel assemblies having a maximum radially averagedU-235 enrichment of 4.80 weight percent;b. keff < 1.0 if fully flooded with unborated water,which includes an allowance for biases anduncertainties as described in Section 9.1 of theUFSAR;c. keff 0.95 if fully flooded with water borated to900 ppm, which includes an allowance for biases anduncertainties as described in Section 9.1 of theUFSAR.d. A nominal 9.5 inch center-to-center distancebetween adjacent storage cell locations.

e. Region 1: Fuel shall be stored in a checkerboard (two-out-of-four) storage pattern.

Fuel thatqualifies to be stored in Regions 1, 2, 3, or 4 inaccordance with Figures 3.7.17-1, 3.7.17-2, or3.7.17-3, may be stored in Region 1.f. Region 2: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2(two-out-of-twelve) assemblies and Region 4(ten-out-of-twelve) assemblies are mixed as shownin Section 9.1 of the UFSAR. Only fuel thatqualifies to be stored in Regions 2, 3, or 4, inaccordance with Figures 3.7.17-1, 3.1.17-2, or3.7.17-3, may be stored in Region 2.g. Region 3: Fuel shall be stored in a four-out-of-four storage pattern.

Only fuel that qualifies tobe stored in Regions 3 or 4, in accordance withFigures 3.7.17-2 or 3.7.17-3, may be stored inRegion 3.(conti nued)PALO VERDE UNITS 1,2,34.0-2PALOVERE UITS1.23 40-2AMENDMENT NO. 47~

I Before SEP transition IDesign Features4.04.0 DESIGN FEATURES (continued)

h. Region 4: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2(two-out-of-twelve) assemblies and Region 4(ten-out-of-twelve) assemblies are mixed as shownin Section 9.1 of the UFSAR. Only fuel thatqualifies to be stored in Region 4 in accordance with Figure 3.7.17-3 shall be stored in Region 4.4.3.1.2 The new fuel storage racks are designed and shall bemaintained with:a. Fuel assemblies having a maximum radially averagedU-235 enrichment of 4.80 weight percent;b. keff 0.95 if fully flooded with unborated water,which includes an allowance for biases anduncertainties as described in Section 9.1 of theUFSAR:c. keff 0.98 if moderated by aqueous foam, whichincludes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR; andd. A nominal 17 inch center to center distance betweenfuel assemblies placed in the storage racks.4.3.2 DrainageThe spent fuel storage pool is designed and shall be maintained toprevent inadvertent draining of the pool below elevation 137 feet -6 inches.4.3.3 CapacityThe spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuelassemblies.

PALO VERDE UNITS 1,2,34.0-3PAO EDEUNT 12, .03AMENDMENT NO. 11 o ...

After SFP transition Design Features4.04.0 DESIGN FEATURES (continued) 4.3 Fuel Storage4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall bemaintained with:46a. Fuel assemblies having/ maximum radially averagedU-235 enrichment .weight percent:b. keff < 1.0 if fully flooded with unborated water.which includes an allowance for biases anduncertainties as described in Section 9.1 of theUFSAR-c. keff -< 0.95 if fully flooded with water borated to~ppm, which includes an allowance for biases andunertainties as described in Section 9.1 of the1460 UFSAR.d. A nominal 9.5 inch center-to-center distancebetween adjacent storage cell locations.

e. R~on !:Fuel shall bc stored in a checkerboa=rd Fuel assemblies are .. ..... --.classified in Fuel Regions qualifie to be tore in Regon 1 2 3, or in1-6 as shown in Tables a ccordance w.i th Figu=res 3.7.17 1, 3.7.17 2, or3.7.17-1 through 3. 7.17 3 .. mayb in Rcgn 1.3.7.17-5. P Fe,,l shall be a; .......tin-3 byIstrg pattern in ..hich Regon 2( .. = atenrutontele seble arfel mixed as. shownaccordance wit FiguQres 3.7.17 1, 3.7/.17%

2,-n orRegionnu3.

PALO VERDE UNITS 1,2,340-AMNETNO

.&4.0-2AMENDMENT NO. 117, 125 After SFP transition Design Featuresi 4.04.0 DESIGN FEATURES (continued) 4.3.1.2 Th enewfuel Fuorag rhalls bre dstored ind sa llcat be3ai bta ine witorgh ateni-hch go(t.oF uteflwlc assemblies hvn m a ndmu ra gionl avrae(t35enrouioctele)t afswembieht aerenmxdt shwincerectiones9.1 dof cthe edUF n S. tion 9.1& of thqualifies to blowne store binaRegio and uccordaincies ait Fiuesrib3 hllb soed in Regtion 9.Ifte FA nm.Ainta ined with inhc, r 465t ene itnc ewea.fuel assemblies hlavin maximumoradill raverageU-235raiagientrichmenteor eih eretThesp bn kefue strg 0.95 if fullyflooed withalunboaed waitaiert preveninawhichn drincldeing allowanepfor biaswesevatind u3 et nchertitessdsrbd nScin91o hThespent includstoane allowanceforsbiase and uncl ermitaintes withstoasgescrpaibyliied iSetiono 9.1ofe then UF3AR andlfuesaseblesplce hisorgeraks PALO VERDE UNITS 1,2,34.0-3PAL VRD UITS1,.34.-3AMENDMENT NO. 1-17,12 ..

IBefore SFP transitionI Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)

4. In Regulatory Guide 1.129, Regulatory Position 3,Subsection 5.4.1, "State of Charge Indicator,"

thefollowing statements in paragraph (d) may be omitted:"When it has been recorded that the charging current hasstabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially highcharging current decreases sharply and the battery voltagerises to approach the charger output voltage."

5. In lieu of RG 1.129. Regulatory Position 7, Subsection 7.6, "Restoration,"

the following may be used: "Following the test, record the float voltage of each cell of thestring."b. The program shall include the following provisions:

1. Actions to restore battery cells with float voltage<2.13 V;2. Actions to determine whether the float voltage of theremaining battery cells is 2.13 V when the floatvoltage of a battery cell has been found to be<2.13 V:3. Actions to equalize and test battery cells that hadbeen discovered with electrolyte level below the topof the plates:4. Limits on average electrolyte temperature, batteryconnection resistance, and battery terminal voltage;and5. A requirement to obtain specific gravity readings ofall cells at each discharge test, consistent withmanufacturer recommendations.

PALO VERDE UNITS 1,2,3 551 MNMN O 95.5-19AMENDMENT NO.

IAfter SFP TransitionI Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)

4. In Regulatory Guide 1.129. Regulatory Position 3,Subsection 5.4.1, "State of Charge Indicator."

thefollowing statements in paragraph (d) may be omitted:"When it has been recorded that the charging current hasstabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially highcharging current decreases sharply and the battery voltagerises to approach the charger output voltage."

5. In lieu of RG 1.129, Regulatory Position 7, Subsection 1.6. "Restoration."

the following may be used: "Following the test, record the float voltage of each cell of thestring."b. The program shall include the following provisions:

1. Actions to restore battery cells with float voltage<2.13 V:2. Actions to determine whether the float voltage of theremaining battery cells is 2.13 V when the floatvoltage of a battery cell has been found to be<2.13 V:3. Actions to equalize and test battery cells that hadbeen discovered with electrolyte level below the topof the plates:4. Limits on average electrolyte temperature, batteryconnection resistance, and battery terminal voltage:and5. A requirement to obtain specific gravity readings ofall cells at each discharge test, consistent withmanufacturer recommendations.

page 5.5-19PALO VERDE UNITS 1,2.3 551 MNMN O 95.5-19AMENDMENT NO.

[After SFP Transition]

Insert for page 5.5-195.5.21 Spent Fuel Storagqe Rack Neutron Absorber Monitoring ProgqramCertain storage cells in the spent fuel storage racks utilize neutron absorbing materialthat is credited in the spent fuel storage rack criticality safety analysis to ensure thelimitations of Technical Specifications 3.7.17 and 4.3.1.1 are maintained.

In order to ensure the reliability of the neutron absorber

material, a monitoring programis provided to confirm the assumptions in the spent fuel pool criticality safety analysis.

The Spent Fuel Storage Rack Neutron Absorber Monitoring Program shall requireperiodic inspection and monitoring of spent fuel pool test coupons and neutron absorberinserts on a performance-based frequency, not to exceed 10 years.Test coupons shall be inspected as part of the monitoring program.

These inspections shall include visual, B-10 areal density and corrosion rate.Visual in-situ inspections of inserts shall also be part of the program to monitor for signsof degradation.

In addition, an insert shall be removed periodically for visual inspection, thickness measurements, and determination of retention force.

Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 2Revised Technical Specifications Pages (Clean Copy)(Pages Provided for Before and After SEP Transition) 3.7.17-13.7.17-23.7.17-33.7.17-43.7.17-53.7.17-63.7.17-74.0-24.0-35.5-195.5-20 Before SFP transition Spent Fuel Assembly Storage3.7.173.7 PLANT SYSTEMS3.7.17 Spent Fuel Assembly StorageLCO 3.7.17The combination of initial enrichment, burnup, and decaytime of each fuel assembly stored in each of the fourregions of the fuel storage pool shall be within theacceptable burnup domain for each region as shown in Figures3.7.17-1, 3.7.17-2, or 3.7.17-3, and described inSpecification 4.3.1.1.APPLICABILITY:

Whenever any fuelpool.assembly is stored in the fuel storageACTIONS ________________

CONDITION REQUIRED ACTION COMPLETION TIMEA. Requirements of the A.1------NOTE----

LCO not met. LCO 3.0.3 is notapplicable.

Initiate action to Immediately move the noncomplying fuel assembly into anappropriate region.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior toenrichment, burnup, and decay time of the storing thefuel assembly is in accordance with Figures fuel assembly3.7.17-1, 3.7.17-2, or 3.7.17-3, and in the fuelSpecification 4.3.1.1.

storage pool.PALO VERDE UNITS 1,2,3 371- MNMN O -~3.7.17-1AMENDMENT NO.

Before SFP transition Spent Fuel Assembly Storage3.7.17Figure 3.7.17-1ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT forRegi on 220000EE1.5 2.0 2.5 3.03.54.0Initial Enrichment, weight %4.5

• 5.04.80%limitingenrichment PALO VERDE UNITS 1,2,3 371- MNMN O3.7.17-2AMENDMENT NO.

Before SFP transition Spent Fuel Assembly Storage3.7.17Figure 3.7.17-2ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT forRegi on 3(at decay times from 0 to 20 years)Eci1.5 2.0 2.5 3.0 3.5 4.0 4.5

• 5.0Initial Enrichment, weight % 4.80%--q Oyers --m-5yars l~ears --X-15ear -- --2yeas ....limitingflmj --Oyars -U-yers ~-1yeas -E-l~yars oyers enrichment DecaylPALO VERDE UNITS 1,2,3 371- MNMN O 23.7.17- 3AMENDMENT NO.

Before SFP transition Spent Fuel Assembly Storage3.7.17Figure 3.7.17-3ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT forRegi on 4(at decay times from 0 to 20 years)5000045000BSA14 + -~ -S'IvuuU ______________

ACCE TABLE fo Region 435000 ____________

30000BBS~.nnn ___________

I-E 20000(I)(0J,I)150001000050000-1.5JSSSSSNOT ACC EPTABLE ~or Region j4/ -U--BSii~-i I SBB___ I ___ ___ ___ ___ __ B-#A~-- 4 + + +/.JSSSBS---S0BSBSSBBcaytime.:

+ t A------N+ote: Assern4Iy eflgbJ for Regi in 4 facua IBU>* U requiefent forjiI ghen irai ech ent and currnt d2-.0 2.5 3.0 3.5 4.0 4.5 a 5.0Initial Enrichment, weight % 4.80%limitinga -U-!--5 years -~-k--l years --UP-15 years --O-20 years .Jenrichment Decaylimej ye.PALO VERDE UNITS 1,2,3 371- MNMN O 23.7.17-4AMENDMENT NO.

After SFP transition ISpent Fuel Assembly Storage3.7.173.7 PLANT SYSTEMS3.7.17 Spent Fuel Assembly StorageLCO 3.7.17The combination of i niti al enrichment, burnup, and decaytime of each fuel assembly shall be in compliance with therequirements specified in Tables 3.7.17-1 through 3.7.17-5.

APPLICABILITY:

Whenever any fuelpool.assembly is stored in the fuel storageACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Requirements of the A.1------NOTE----

LCO not met. LCO 3.0.3 is notapplicable.

Initiate action to Immediately move the noncomplying fuel assembly into anappropriate region.SURVEILLANCEREQUIREMENTS__________

SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the initial Prior toenrichment, burnup, and decay time of the storing thefuel assembly is in accordance with Tables fuel assembly3.7.17-1 through 3.7.17-5, Figure 3.7.17-1, in the fueland Specification 4.3.1.1.

storage pool.PALO VERDE UNITS 1,2,337171AEDNT O.I,3.7.17-1AMENDMENT NO.

After SFP transition Spent Fuel Assembly Storage3.7.17Table 3.7.17-1Fuel RegionsRanked by Reactivity Fuel Region 1 Highest Reactivity (See Note 2)Fuel Region 2Fuel Region 3Fuel Region 4Fuel Region 5Fuel Region 6 Lowest Reactivity Notes:1. Fuel Regions are defined by assembly average burnup. initial enrichment' and decaytime as provided by Table 3.7.11-2 through Table 3.7.17-5.

2. Fuel Regions are ranked in order of decreasing reactivity, e.g.. Fuel Region 2 isless reactive than Fuel Region 1. etc.3. Fuel Region 1 contains fuel with an initial maximum radially averaged enrichment upto 4.65 wt% 235U. No burnup is required.

4. Fuel Region 2 contains fuel with an initial maximum radially averaged enrichment upto 4.65 wt% 235U with at least 16.0 GWd/MTU of burnup.5. Fuel Regions 3 through 6 are determined from the minimum burnup (BU) equation andcoefficients provided in Tables 3.7.17-2 through 3.7.17-5.

6. Assembly storage is controlled through the storage arrays defined in Figure 3.7.17-1.

7. Each storage cell in an array can only be populated with assemblies of the FuelRegion defined in the array definition or a lower reactivity Fuel Region.'Initial Enrichment is the nominal 235U enrichment of the central zone region of fuel, excluding axialblankets, prior to reduction in 2350 content due to fuel depletion.

If the fuel assembly contains axialregions of different 235U enrichment values, such as axial blankets, the maximum initial enrichment value is to be utilized.

PALO VERDE UNITS 1,2,337.72AEDNTO.1, 3.7.17-2AMENDMENT NO.

After SEP transition ISpent Fuel Assembly Storage3.7.17Table 3.7.17-2Fue] Region 3: Burnup Requi rement Coefficients Decay Coefficients Time (yr.) Ai NmA A40 -1.5473 15.5395 -39.0197 24.11215 -1.4149 13.9760 -33.6287 18.336910 -1.3012 12.6854 -29.2539 13.687915 -1.0850 10.4694 -22.1380 6.367320 -0.9568 9.1487 -17.9045 2.0337Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

Forinstance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meet the requirements of a FuelRegion, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given bythe curve fit for the assembly "decay 'time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated fromthe following equation:

BU = Ai

• En3 + A2

• En2 + A3

• En + A42. Initial enrichment, En, is the maximum radial average 23enrichment.

Any Envalue between 2.55 wt% 235U and 4.65 wt% 235U may be used. Burnup credit is notrequired for an En below 2.55 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based ondecay time.4. The 20-year coefficients must be used to calculate the minimum BU for anassembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN O 23.7.17-3AMENDMENT NO.

After SFP transition Spent Fuel Assembly Storage3.7.17Table 3.7.17 -3Fue] Region 4: Burnup Requirement Coefficients Decay Coefficients Time (yr.) Ai #0 0.4260 -6.2766 40.9264 -54.68135 0.2333 -4.1545 32.9080 -46.116110 0.4257 -6.2064 39.0371 -51.588915 0.5315 -7.3777 42.5706 -54.752420 0.5222 -7.3897 42.6587 -54.8201Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

Forinstance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meet the requirements of a FuelRegion, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given bythe curve fit for the assembly "decay time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated fromthe following equation:

BU =Ai

• En3 + A2

• En2 + A3

• En + A42. Initial enrichment, En, is the maximum radial average 235U enrichment.

Any Envalue between 1.15 wt% 235U and 4.65 wt% 23may be used. Burnup credit is notrequired for an En below 1.75 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based ondecay time.4. The 20-year coefficients must be used to calculate the minimum BU for anassembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN O 23.7.17-4AMENDMENT NO.

After SFP transition ISpent Fuel Assembly Storage3.7.17Table 3.7.17 -4Fue] Region 5: Burnup Requirement Coefficients Decay Coeffi ci entsTime (yr.) Ai km# A40 -0.1114 -0.4230 20.9136 -32.85515 -0. 1232 -0. 4463 20. 8337 -32. 606810 -0. 2357 0.4892 18. 0192 -30. 004215 -0.1402 -0. 4523 20. 3745 -31. 756520 -0. 0999 -0. 8152 21. 0059 -31. 9911Notes :1. Relevant uncertainties are explicitly included in the criticality analysis.

Forinstance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meet the requirements of a FuelRegion, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given bythe curve fit for the assembly "decay time" and "initial enrichment."

Thespeci fi c minimum burnup (BU) requi red for each fuel assembly is calculated fromthe fol lowing equati on:BU= Ai

• En3 + A2

• En2 + A3

• En + A42. Initial enrichment, En. is the maximum radial average 235U enrichment.

Any Envalue between 1.65 wt% 235U and 4.65 wt% 235U may be used. Burnup credit is notrequired for an En below 1.65 wt% 235U.3. It is acceptable to linearly interpolate between calculated BU limits based ondecay time.4. The 20-year coefficients must be used to calculate the minimum BU for anassembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN D3.7.17-5AMENDMENT NO.

After SFP transition Spent Fuel Assembly Storage3.7.17Table 3.7.17 -5Fuel Region 6: Burnup Requirement Coefficients Decay Coefficients Time (yr.) A A2 A4 A0 0.7732 -9.3583 49.6577 -54.68475 0.7117 -8.4920 45.1124 -49.728210 0.6002 -7.2638 40.2603 -44.934815 0.5027 -6.2842 36.6715 -41.493420 0.2483 -3.7639 28.8269 -34.6419Notes:1. Relevant uncertainties are explicitly included in the criticality analysis.

Forinstance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meet the requirements of a FuelRegion, the assembly burnup must exceed the "minimum burnup" (GWd/MTU) given bythe curve fit for the assembly "decay time" and "initial enrichment."

Thespecific minimum burnup (BU) required for each fuel assembly is calculated fromthe fol lowi ng equati on:BU =A1

• En3 + Am

• En2 + A3

• En + A42. Initial enrichment, En. is the maximum radial average 235U enrichment.

Any Envalue between 1.45 wt% 235U and 4.65 wt% 235U may be used. Burnup credit is notrequired for an En below 1.45 wt% 2350.3. It is acceptable to linearly interpolate between calculated BU limits based ondecay time.4. The 20-year coefficients must be used to calculate the minimum BU for anassembly with a decay time of greater than 20 years.PALO VERDE UNITS 1,2,3 371- MNMN O3.7.17-6AMENDMENT NO.

After SFP transition Spent Fuel Assembly Storage3.7.17Figure 3.7.17-1Allowable Storage ArraysArray A 1 XTwo Region 1 assemblies (1) checkerboarded with two blocked cells (X).The Region 1 assemblies are each in a cell with a stainless steelL-insert.

No NETCO-SNAP-IN inserts are credited.X 1Array B 1 TCTwo Region 1 assemblies (1) checkerboarded with two cells containing trash cans (TC). The Region 1 assemblies are each in a cell with astainless steel L-insert.

Every cell without a stainless steel L- TC Iinsert must contain a NETCO-SNAP-IN insert.Array CTwo Region 2 assemblies (2) checkerboarded with one Region 3 assembly 2 X(3) and one blocked cell (X). The Region 2 assemblies are each in a ....... ..............

cell with a stainless steel L-insert.

The Region 3 assembly is in a 3 2cell containing a NETCO-SNAP-IN insert.Array 0One Region 2 assembly (2) checkerboarded with three Region 4 2 4assemblies (4). The Region 2 assembly and the diagonally located ..Region 4 assembly are each in a storage cell with a stainless steel L-insert. The two storage cells without a stainless steel L-insert 4 4contain a NETCO-SNAP-IN insert.Array E 5 5Four Region 5 assemblies (5). Two storage cells contain a stainless steel L-insert.

One cell contains a NETCO-SNAP-IN insert. One5storage cell contains no insert.5 5Array F 6 6Four Region 6 assemblies (6). Two storage cells contain a stainless steel L-insert.

The other two cells contain no inserts.

6 6Notes-1. The shaded locations indicate cells which contain a stainless steel [-insert.

2. A blocked cell (X) contains a blocking device and only water in the active fuel region.3. NETCO-SNAP-IN inserts must be oriented in the same direction as the stainless steel [-inserts.

4. NETCO-SNAP-IN inserts are only located in cells without a stainless steel [-insert.

5. Any cell containing a fuel assembly or a TC may instead be an empty (water-filled) cell inall storage arrays.6.Any storage array location designated for a fuel assembly may be replaced with non-fissile material.

PALO VERDE UNITS 1,2,3 371- MNMN O3.7.17-7AMENDMENT NO.

Before SFP transition Design Features4.04.0 DESIGN FEATURES (continued) 4.3 Fuel Storage4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall bemaintained with:a. Fuel assemblies having a maximum radially averagedU-235 enrichment of 4.80 weight percent;b. keff < 1.0 if fully flooded with unborated water,which includes an allowance for biases anduncertainties as descri bed i n Secti on 9.1I of theUFSAR;c. keff -< 0.-95 if fully flooded with water borated to900 ppm, which includes an allowance for biases anduncertainties as described in Section 9.1 of theUFSAR.d. A nominal 9.5 inch center-to-center distancebetween adjacent storage cell locations.

e. Region 1: Fuel shall be stored in a checkerboard (two-out-of-four) storage pattern.

Fuel thatqualifies to be stored in Regions 1, 2, 3, or 4 inaccordance with Figures 3.7.17-1, 3.7.17-2, or3.7.17 -3, may be stored in Region 1.f. Region 2: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2(two-out-of-twelve) assemblies and Region 4(ten-out-of-twelve) assemblies are mixed as shownin Section 9.1 of the UFSAR. Only fuel thatqualifies to be stored in Regions 2, 3, or 4, inaccordance with Figures 3.7.17-1, 3.7.17-2, or3.7.17-3, may be stored in Region 2.g. Region 3: Fuel shall be stored in a four-out-of-four storage pattern.

Only fuel that qualifies tobe stored in Regions 3 or 4, in accordance withFigures 3.7.17-2 or 3.7.17-3, may be stored inRegi on 3.(conti nued)PALO VERDE UNITS 1,2,3 402AEDETN.1k 4.0-2AMENDMENT NO.

Before SEP transition Design Features4.04.0 DESIGN FEATURES (continued)

h. Region 4: Fuel shall be stored in a repeating 3-by-4 storage pattern in which Region 2(two-out-of-twelve) assemblies and Region 4(ten-out-of-twelve) assemblies are mixed as shownin Section 9.1 of the UFSAR. Only fuel thatqualifies to be stored in Region 4 in accordance with Figure 3.7.17-3 shall be stored in Region 4.4.3.1.2 The new fuel storage racks are designed and shall bemaintained with:a. Fuel assemblies having a maximum radially averagedU-235 enrichment of 4.80 weight percent;b. keff 0.95 if fully flooded with unborated water,which includes an allowance for biases anduncertainties as described in Section 9.1 of theUFSAR:c. keff 0.98 if moderated by aqueous foam, whichincludes an allowance for biases and uncertainties as described in Section 9.1 of the UFSAR; andd. A nominal 17 inch center to center distance betweenfuel assemblies placed in the storage racks.,.4.3.2 DrainageThe spent fuel storage pool is designed and shall be maintained toprevent inadvertent draining of the pool below elevation 137 feet -6 inches.4.3.3 CapacityThe spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuelassemblies.

PALO VERDE UNITS 1,2,3 403AEDETN.I~

4.0-3AMENDMENT NO.

After SFP transition Design Features4.04.0 k[DESIGN FEATURES(conti nued)4.3 Fuel Storage4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall bemai ntai ned with :a. Fuel assemblies having a maximum radially averagedU-235 enrichment of 4.65 weight percent;b. keff < 1.0 if fully flooded with unborated water,which includes an allowance for biases anduncertai nti es as descri bed in Secti on 9.1 of theUFSAR;c. keff -< 0.95 if fully flooded with water borated to1460 ppm, which includes an allowance for biasesand uncertainties as described in Section 9.1 ofthe UFSAR.d. A nominal 9.5 inch center-to-center distancebetween adjacent storage cell locations.

e. Fuel assemblies, are classified in Fuel Regions 1-6as shown in Tables 3.7.17-1 through 3.7.17-5.

(conti nued)PALO VERDE UNITS 1,2,340-AMNETNO 4.0-2AMENDMENT NO.

After SFP transition Design Features4.04.0 DESIGN FEATURES (continued) 4.3.1.2 The new fuel storage racks are designed and shall bemaintained with:a. Fuel assemblies having a maximum radially averagedU-235 enrichment of 4.65 weight percent;b. keff 0.95 if fully flooded with unborated water,which includes an allowance for biases anduncertai nti es as descri bed i n Secti on 9.1I of theU FSAR;c. keff 0.98 if moderated by aqueous foam, whichi ncl udes an all owance for bi ases and uncertai nti esas described in Section 9.1 of the UFSAR; andd. A nominal 18 inch (east-west) and 31 inch (north-south) center-to-center distance between fuelassemblies placed in the storage racks.4.3.2 DrainageThe spent fuel storage pool is designed and shall be maintained toprevent inadvertent draining of the pool below elevation 137 feet -6 inches.4.3.3 CapacityThe spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuelassemblies.

PALO VERDE UNITS 1,2,340-AMNETNO 2g4.0-3AMENDMENT NO.

Before SFP transition Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)

4. In Regulatory Guide 1.129, Regulatory Position 3,Subsection 5.4.1, "State of Charge Indicator,"

thefollowing statements in paragraph (d) may be omitted:"When it has been recorded that the charging current hasstabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially highcharging current decreases sharply and the battery voltagerises to approach the charger output voltage."

5. In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration,"

the following may be used: "Following the test, record the float voltage of each cell of thestring."b. The program shall include the following provisions:

1. Actions to restore battery cells with float voltage<2.13 V;2. Actions to determine whether the float voltage of theremaining battery cells is 2.13 V when the floatvoltage of a battery cell has been found to be<2.13 V;3. Actions to equalize and test battery cells that hadbeen discovered with electrolyte level below the topof the plates;4. Limits on average electrolyte temperature, batteryconnection resistance, and battery terminal voltage;and5. A requirement to obtain specific gravity readings ofall cells at each discharge test, consistent withmanufacturer recommendations.

PALO VERDE UNITS 1,2,3 551 MNMN O 95.5-19AMENDMENT NO.

After SFP transition Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.19 Battery Monitoring and Maintenance Program (continued)

4. In Regulatory Guide 1.129, Regulatory Position 3,Subsection 5.4.1, "State of Charge Indicator,"

thefollowing statements in paragraph (d) may be omitted:"When it has been recorded that the charging current hasstabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge.These measurements shall be made after the initially highcharging current decreases sharply and the battery voltagerises to approach the charger output voltage."

5. In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration,"

the following may be used: "Following the test, record the float voltage of each cell of thestring."b. The program shall include the following provisions:

1. Actions to restore battery cells with float voltage<2.13 V;2. Actions to determine whether the float voltage of theremaining battery cells is 2.13 V when the floatvoltage of a battery cell has been found to be<2.13 V;3. Actions to equalize and test battery cells that hadbeen discovered with electrolyte level below the topof the plates;4. Limits on average electrolyte temperature, batteryconnection resistance, and battery terminal voltage;and5. A requirement to obtain specific gravity readings ofall cells at each discharge test, consistent withmanufacturer recommendations.

PALO VERDE UNITS 1,2,35519AEDNT O.495.5-19AMENDMENT NO.

After SFP transition Programs and Manuals5.55.5 Programs and Manuals (continued) 5.5.21 Spent Fuel Storage Rack Neutron Absorber Monitoring ProgramCertain storage cells in the spent fuel storage racks utilizeneutron absorbing material that is credited in the spent fuelstorage rack criticality safety analysis to ensure thelimitations of Technical Specifications 3.7.17 and 4.3.1.1 aremaintained.

In order to ensure the reliability of the neutron absorbermaterial, a monitoring program is provided to confirm theassumptions in the spent fuel pool criticality safety analysis.

The Spent Fuel Storage Rack Neutron Absorber Monitoring Programshall require periodic inspection and monitoring of spent fuelpool test coupons and neutron absorber inserts on aperformance-based frequency, not to exceed 10 years.Test coupons shall be inspected as part of the monitoring program.

These inspections shall include visual, B-IO arealdensity and corrosion rate.Visual in-situ inspections of inserts shall also be part of theprogram to monitor for signs of degradation.

In addition, aninsert shall be removed periodically for visual inspection, thickness measurements, and determination of retention force.PALO VERDE UNITS 1,2,35520AEDNT O.195.5-20AMENDMENT NO.

Enclosure Description and Assessment of Proposed License Amendment ATTACHM ENT 3Marked-up Technical Specifications Bases Pages(Pages Provided for Before and After SEP Transition)

B 3.7.15-1B 3.7.15-2B 3.7.17-1B 3.7.17-2B 3.7.17-3B 3.7.17-4B 3.7.17-5B 3.7.17-6C IBefore SFP transitionJ Fuel Storage Pool Boron Concentration B 3.7.15B 3.7 PLANT SYSTEMSB 3.7.15 Fuel Storage Pool Boron Concentration BASESBACKGROUND As described in LCO 3.7.17, 'Spent Fuel Assembly Storage,"

fuel assemblies are stored in the spent fuel racks inaccordance with criteria based on initial enrichment anddischarge burnup. Although the water in the spent fuel poolis normally borated to >_ 2150 ppm, the criteria that limitthe storage of a fuel assembly to specific rack locations isconservatively developed without taking credit for boron.In order to maintain the spent fuel pool keff < 1.0, asoluble boron concentration of 900 ppm is required tomaintain the spent fuel pool keff _< 0.95 assuming the mostlimiting single fuel mishandling accident.

APPLICABLE SAFETY ANALYSESA fuel assembly could be inadvertently loaded into a spentfuel rack location not allowed by LCO 3.7.17 (e.g., anunirradiated fuel assembly or an insufficiently depletedfuel assembly).

Another type of postulated accident isassociated with a fuel assembly that is dropped onto thefully loaded fuel pool storage rack or between a rack andthe pool walls. These incidents could have a positivereactivity effect, decreasing the margin to criticality.

However, the negative reactivity effect of the soluble boroncompensates for the increased reactivity caused by thesepostulated accident scenarios.

The concentration of dissolved boron in the fuel poolsatisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCOThe specified concentration of dissolved boron in the fuelpool preserves the assumptions used in the analyses of thepotential accident scenarios described above. Thisconcentration of dissolved boron is the minimum requiredconcentration for fuel assembly storage and movement withinthe fuel pool.APPLICABILITY This LCO applies whenever any fuel assembly is stored inthe spent fuel pool in order to comply with theTS 4.3.1.1.c design requirement that keff 0.95.(conti nued)PALO VERDE UNITS 1,2,3B37.51RVSO B 3.7.15-1REVISION 3

Before SFP transition Fuel Storage Pool Boron Concentration B 3.7.15BASES (continued)

ACTIONSA.1 and A.2The Required Actions are modified by a Note indicating thatLCO 3.0.3 does not apply.When the concentration of boron in the spent fuel pooi isless than required, immediate action must be taken topreclude an accident from happening or to mitigate theconsequences of an accident in progress.

This is mostefficiently achieved by immediately suspending the movementof fuel assemblies.

This does not preclude the movement offuel assemblies to a safe position.

In addition, actionmust be immediately initiated to restore boron concentration to within limit.If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3would not specify any action. If moving fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement isindependent of reactor operation.

Therefore, inability tosuspend movement of fuel assemblies is not sufficient reasonto require a reactor shutdown.

SURVEILLANCE SR 3.7.15.1REQUI REMENTSThis SR verifies that the concentration of boron in thespent fuel pool is within the required limit. As long asthis SR is met, the analyzed incidents are fully addressed.

The Surveillance Frequency is controlled under theSurveillance Frequency Control Program.REFERENCES

1. UFSAR, Section 9.1.2.2. PVNGS Operating License Amendments 82, 69 and 54 forUnits 1, 2 and 3, respectively, and associated NRCSafety Evaluation dated September 30, 1994.3. 13-N-001-1900-1221-1, "Palo Verde Spent Fuel PoolCriticality Analysis,"

ABB calculation A-PV-FE-0106, revision 3, dated January 15, 1999.PALO VERDE UNITS 1,2,3B37152RVSO

.B 3.7.15-2REVISION Ifer FP ransition Fuel Storage Pool Boron Concentration B 3.7.15B 3.7 PLANT SYSTEMSB 3.7.15 Fuel Storage Pool Boron Concentration

,an(time.BASESBACKGROUND As dscrbedin.CO 3.7.17. "Spent Fuel Assembly Afuel decieassembliesi tre stored in the spent fuel racks accordance with ,4iteria based on initial enrichment

~discharge burnup' Although the .ater .pent fue pool is .. nomly borte to 2l50 ppm the critei that limtthe strg of.. a fue to sp.cifi rac location-s, isoluble boron concentration of m is required tomaintain the spent fuel pool kef _< 0.ming the mostlimiting fuel mishandling accident. 1460APPLICABLE SAFETY ANALYSESThere could also bea misload ofmultiple fuelassemblies into fuelrack locations notallowed by LCO3.7.17.A fuel assembly could be inadvertently loaded into a spentfuel rack location not allowed by LCO 3.7.17 (e.g., anunirradiated fuel assembly or an insufficiently depletedfuel Another type of postulated accident is a fuel assembly that is dropped onto thefully 1 ~ted fuel pool storage rack or between a rack andthe 4mol walls. These incidents could have a positiveactivity effect, decreasing the margin to criticality.

However, the negative reactivity effect of the soluble boroncompensates for the increased reactivity caused by thesepostulated accident scenarios.

The concentration of dissolved boron in the fuel poolsatisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCOThe specified concentration of dissolved boron in the fuelpool preserves the assumptions used in the analyses of thepotential accident scenarios described above. Thisconcentration of dissolved boron is the minimum requiredconcentration for fuel assembly storage and movement withinthe fuel pool.APPLICABILITY This LCO applies whenever any fuel assembly is stored inthe spent fuel pool in order to comply with theTS 4.3.1.1.c design requirement that keff 0.95.(conti nued)PALO VERDE UNITS 1,2,3B37151RVSO B 3.7.15-1REVISION IAfter SFP transition Fuel Storage Pool Boron Concentration B 3.7.15BASES (continued)

ACTIONSA.1 and A.2The Required Actions are modified by a Note indicating thatLCO 3.0.3 does not apply.When the concentration of boron in the spent fuel pool isless than required, immediate action must be taken topreclude an accident from happening or to mitigate theconsequences of an accident in progress.

This is mostefficiently achieved by immediately suspending the movementof fuel assemblies.

This does not preclude the movement offuel assemblies to a safe position.

In addition, actionmust be immediately initiated to restore boron concentration to within limit.If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3would not specify any action. If moving fuel assemblies while in MODE 1, 2, 3. or 4, the fuel movement isindependent of reactor operation.

Therefore, inability tosuspend movement of fuel assemblies is not sufficient reasonto require a reactor shutdown.

SURVEILLANCE SR 3.7.15.1REQUIREMENTS This SR verifies that the concentration of boron in thespent fuel pool is within the required limit. As long asthis SR is met, the analyzed incidents are fully addressed.

The Surveillance Frequency is controlled under theSurveillance Frequency Control Program.REFERENCES

1. UFSAR, Section 9.1.2.2. PVNGS Opcr,:ting ic-c-nse Amcndmcnts

82. 69 3nd{ 51 fo-'r""Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3" (Proprietary),

WCAP-1 8030-P, Revision 0,September 2015.PALO VERDE UNITS 1,2,3B 3.7.15-2REVISION I Before SFP transition ISpent Fuel Assembly StorageB 3.7.17B 3.7 PLANT SYSTEMSB 3.7.17 Spent Fuel Assembly StorageBAS ESBAGCKGROUND The spent fuel storage is designed to store either new(nonirradiated) nuclear fuel assemblies, or burned(irradiated) fuel assemblies in a vertical configuration underwater.

The storage pool was originally designed to storeup to 1329 fuel assemblies in a borated fuel storage mode.The current storage configuration, which allows credit to betaken for boron concentrati.on, burnup, and decay time, anddoes not require neutron absorbing (boraflex) storage cans,provides for a maximum storage of 1209 fuel assemblies in afour-region configuration.

The design basis of the spent fuelcooling system, however, is to provide adequate cooling to thespent fuel during all operating conditions (including fullcore offload) for only 1205 fuel assemblies (UFSAR section9.1.3). Therefore, an additional four spaces are mechanically blocked to limit the maximum number of fuel assemblies thatmay be stored in the spent fuel storage pool to 1205.Region 1 is comprised of two 9x8 storage racks and one 12x8storage rack. Cell blocking devices are placed in every otherstorage cell location in Region 1 to maintain a two-out-of-four checkerboard configuration.

These cell blocking devicesprevent inadvertent insertion of a fuel assembly into a cellthat is not allowed to contain a fuel assembly.

Region 3 is comprised of three 9x8 storage racks and one 9x9storage rack in Units 2 and 3. Region 3 is comprised of four9x8 storage racks and one 9x9 storage rack in Unit 1. Sincefuel assemblies may be stored in every Region 3 cell location, no cell blocking devices are installed in Region 3.Regions 2 and 4 are mixed and are comprised of seven 9x8storage racks and three 12x8 storage racks in Units 2 and 3,Regions 2 and 4 are mixed and are comprised of six 9x8 storageracks and three 12x8 storage racks in Unit 1. Regions 2 and 4are mixed in a repeating 3x4 storage pattern in which two-out-of-twelve cell locations are designated Region 2 and ten-out-of-twelve cell locations are designated Region 4 (see UFSARFigures 9.1-7 and 9.1-7A).

Since fuel assemblies may bestored in every Region 2 and Region 4 cell location, no cellblocking devices are installed in Region 2 and Region 4.(conti nued)PALO VERDE UNITS 1,2,3B371-1RVSO B 3.7.17-iREVISION

[Before SFP transition Spent Fuel Assembly StorageB 3.7.17BASESBACKGROUND (conti nued)The spent fuel storage cells are installed in parallel rowswith a nominal center-to-center spacing of 9.5 inches. Thisspacing, a minimum soluble boron concentration of 900 ppm,and the storage of fuel in the appropriate region based onassembly burnup in accordance with TS Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3 is sufficient to maintain a keff of0.95 for fuel of original maximum radially averagedenrichment of up to 4.80%.APPLICABLE SAFETY ANALYSESThe spent fuel storage pool is designed for non-criticality by use of adequate

spacing, credit for boronconcentration, and the storage of fuel in the appropriate region based on assembly burnup in accordance withTS Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3.

The designrequirements related to criticality (TS 4.3.1.1) arekeff < 1.0 assuming no credit for boron and keff 0.95taking credit for soluble boron. The burnup versusenrichment requirements (TS Figures 3.7.17-1, 3.7.17-2, and3.7.17-3) are developed assuming keff < 1.0 with no credittaken for soluble boron, and that keff 0.95 assuming asoluble boron concentration of 900 ppm and the mostlimiting single fuel mishandling accident.

The analysis of the reactivity effects of fuel storage inthe spent fuel storage racks was performed by ABB-Combustion Engineering (CE) using the three-dimensional Monte Carlo codeKENO-VA with the updated 44 group ENDF/B-5 neutron crosssection library.

The KENO code has been previously used byCE for the analysis of fuel rack reactivity and have beenbenchmarked against results from numerous criticalexperiments.

These experiments simulate the PVNGS fuelstorage racks as realistically as possible with respect toparameters important to reactivity such as enrichment andassembly spacing.The modeling of Regions 2, 3, and 4 included severalconservative assumptions.

These assumptions neglected thereactivity effects of poison shims in the assemblies andstructural grids. These assumptions tend to increase thecalculated effective multiplication factor (keff) of theracks. The stored fuel assemblies were modeled as CE 16x16assemblies with a nominal pitch of 0.5065 inches between fuelrods, a fuel pellet diameter of 0.3255 inches, and a UO(2)density of 10.31 g/cc.(conti nued)PALO VERDE UNITS 1,2,3B37172RVSO 3B 3.7.17-2REVISION 3

IBefore SFP transition]

Spent Fuel Assembly StorageB 3.7.17BASESAPPLICABLE SAFETY ANALYSES(conti nued)KENO-Va calculations were used to construct curves of burnupversus initial enrichment for decay times in 5 yearincrements from 0 to 20 years for both Regions 3 and 4(TS Figures 3.7.17-2 and 3.7.11-3) such that all points onthe curves produce a keff value (including all biases anduncertainties) of < 1.0 for unborated water. Biasesassociated with methodology and water temperature wereincluded, and uncertainties associated with methodology, KENO-Va calculation, fuel enrichment, fuel rack pitch, fuelrack and L-insert thickness, pellet stack density, andasymmetric fuel assembly loading were included.

KENO-Vacalculations were also performed to determine the solubleboron concentration required to maintain the spent fuel poolkeff (including all biases and uncertainties) 0.95 at a95% probability/95%

confidence level. A soluble boronconcentration of 900 ppm is required to assure that the spentfuel pool keff remains 0.95 at all times. This solubleboron concentration accounts for the positive reactivity effects of the most limiting single fuel mishandling eventand uncertainties associated with fuel assembly reactivity and burnup. This method of reactivity equivalencing has beenaccepted by the NRC (Reference

3) and used for numerous otherspent fuel storage pools that take credit for burnup, decaytime, and soluble boron.Most abnormal storage conditions will not result in anincrease in the keff of the racks. However, it is possibleto postulate events, with a burnup and enrichment combination outside of the acceptable area in TS Figure 3.1.17-1, or witha burnup, decay time, and enrichment combination outside ofthe acceptable area in IS Figures 3.7.17-2 or 3.7.17-3, whichcould lead to an increase in reactivity.

These events wouldinclude an assembly drop on top of a rack or between a rackand the pool walls, or the misloading of an assembly.

Forsuch events, partial credit may be taken for the solubleboron in the spent fuel pool water to ensure protection against a criticality accident since the staff does notrequire the assumption of two unlikely, independent, concurrent events (double contingency principle).

Although asoluble boron concentration of only 900 ppm is required toassure that keff remains 0.95 assuming the single mostlimiting fuel mishandling event, TS 3.7.15 conservatively requires the presence of 2150 ppm of soluble boron in thespent fuel pool water. As such, the reduction in keff causedby the required soluble boron concentration more than offsetsthe reactivity addition caused by credible accidents, and thestaff criterion of keff 0.95 is met at all times.(conti nued)PALO VERDE UNITS 1,2,3B37.-3RVSO 3B 3.7.17-3REVISION 3

IBefore SEP transitionI Spent Fuel Assembly StorageB 3.7.17BASESAPPLICABLE The criticality aspects of the spent fuel pool meet theSAFETY ANALYSES requirements of General Design Criterion 62 for the(continued) prevention of criticality in fuel storage and handling.

The spent fuel pool heat load calculations were based on afull pool with 1205 fuel assemblies.

From the spent fuelpool criticality

analysis, the number of fuel assemblies that can be stored in the four-region configuration is1209 fuel assemblies.

The design basis of the spent fuelcooling system, however, is to provide adequate cooling tothe spent fuel during all operating conditions (including full core offload) for only 1205 fuel assemblies (UFSAR section 9.1.3). Therefore, an additional four spacesare mechanically blocked to limit the maximum number of fuelassemblies that may be stored in the spent fuel storage poolto 1205.The original licensing basis for the spent fuel pool allowedfor spent fuel to be loaded in either a 4x4 array or acheckerboard array, depending on the use of borated poison.A fuel handling accident was assumed to occur with maximumloading of the pool. The fuel pool rack construction precludes more than one assembly from being impacted in afuel handling accident.

The UFSAR analysis conclusion regarding the worst scenario for a dropped assembly (in whichthe horizontal impact of a fuel assembly on top of the Spentfuel assembly damages fuel rods in the dropped assembly butdoes not impact fuel in the stored assemblies) continues tobe limiting.

The spent fuel assembly storage satisfies Criterion 2 of 10CFR 50.36 (c)(2)(ii).

LCO The restrictions on the placement of fuel assemblies withinthe spent fuel pool, according to Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3 in the accompanying LCO, ensures that the keff ofthe spent fuel pool will always remain < 1.0 assuming thepool to be flooded with unborated water. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool according to Figures 3.1.17-1, 3.1.17-2, and 3.7.17-3 in the accompanying LCO.Specification 4.3.1.1 provides additional details for fuelstorage in each of the four Regions.(conti nued)PALO VERDE UNITS 1,2,3B371-4RVSO B 3.7.17-4REVISION IBefore SFP transition Spent Fuel Assembly StorageB 3.7.17BASESAPPLICABILITY This LCO applies whenever any fuel assembly is stored in thespent'fuel pool.ACTIONS A. 1Requi red Acti on A.1 i is modi fi ed by a Note i ndi cating thatLCO 3.0.3 does not apply.When the configuration of fuel assemblies stored in the spentfuel pool is not in accordance with Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3, immediate action must be taken tomake the necessary fuel assembly movement(s) to bring theconfiguration into compliance with Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3.

If moving irradiated fuel assemblies while in MODE 5 or 6,LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuelmovement is independent of reactor operation.

Therefore, ineither case, inability to move fuel assemblies is notsufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.17.1REQU IREMENTSThis SR verifies by administrative means that the initialenrichment and burnup of the fuel assembly is in accordance with Figures 3.7.17-1, 3.7.17-2, and 3.7.17-3 in theaccompanying LCO and Speci fi cati on 4.3.1.1.To manually determine the allowed SFP region for a fuelassembly, the actual burnup is compared to the burnuprequirement for the given initial enrichment andappropriate decay time from Figure 3.7.17-1, 3.7.17-2, or3.7.17-3.

If the actual burnup is greater than or equal tothe burnup requirement, then the fuel assembly is eligibleto be stored in the corresponding region. If the actualburnup is less than the burnup requirement, then thecomparison needs to be repeated using another curve for alower numbered region. Note the fol lowing :(continued)

PALO VERDE UNITS 1,2,3B371-5RVSO 3B 3.7.17-5REVISION 3

SBefore SFP transition ISpent Fuel Assembly StorageB 3.7.17BASESSURVEILLANCE

• that a fuel assembly that does not meet the burnupREQUIREMENTS requirement for Region, 2 must be stored in Region 1,(conti nued)* that any fuel assembly may be stored in Region 1,* that any fuel assembly may be stored in a lower numberedregion than the region for which it qualifies becauseburnup requirements decrease as region numbers decrease(refer also to Tech Spec 4.3.1.1),

• and that comparing actual burnup to the burnuprequirement for zero decay time will always be corrector conservative.

REFERENCES

1. UFSAR, Sections 9.1.2 and 9.1.3.2. PVNGS Operating License Amendments 82, 69, and 54 forUnits 1, 2, and 3 respectively, and associated NRCSafety Evaluation, dated September 30, 1994.3. Letter to T. E. Collins, U.S. NRC to T. Greene, WOG,"Acceptance for Referencing of Licensing TopicalReport WCAP-14416-P, Westinghouse Spent Fuel RackMethodology (TAC NO. M93254)",

October 25, 1996.4. 13-N-001-1900-1221-1, "Palo Verde Spent Fuel PoolCriticality Analysis,"

ABB calculation A-PV-FE-0106, revision 03, dated January 15, 1999.5. Westinghouse letter NF-APS-10-19, "Criticality SafetyEvaluation of the Spent Fuel Pool Map with a ProposedRegion 3 Increase,"

dated February 25, 2010.PALO VERDE UNITS 1,2,3 B371- EIIN~B 3.7.17-6REVISION IAfter SFP transition Spent Fuel Assembly StorageB 3.7.17* The design basis of the spent fuel pool cooling system is toB 3.7 PLANT SYSTEMS IProvide adequate cooling to the spent fuel pool during allIoperating conditions (including full core offload) for up to 1205B 3.7.17 Spent Fuel Assembl lfuel assemblies (UFSAR Section 9.1.3).BASES /I l tBACKGROUND The spent fuel is designed ostor2/'2h4 new(nonirradiated) fuel asse ~lies,--g¢ burned(irradiated) fuel assemblies in Jvertical configuration underwater.

The storage pool sorgnally designed to storeup to 1329 fuel assemblies i eae ue trg oeThe current ,storg config..

ration..

w4. ..hich allow credit. betaken for boron.conentration burnup,..

an dea time, and..dones mnot requirelm mneutron absorbing (boraflex)v ca-nsprovides-ltc a maximu~m stor=ag of 1209Q "fuel assemblies¢ in afour reio configuration=.

.The desig basis of the spent fue,,lcooling syste,-,

however, is to provide adequatea cooling1 to thecore offload frl on nly 1205 asseamblie (UFSTAR sct,,'io~n a 1-- -'- Therefore, an =additional four.spaces.are.mechanically..

bmlockedn to limitf t-he mai~m~um nulmberA t-hatCell de,,'ie are..pnlaceda in ever... otherstoragea cell location in Region 1 to mainti-mn a out- offour checkerbar configuration.

blockin devicespreventf inadverte4-nt-insertion ofc a fueal into a cellthat i not ..alloe to contain a fuel assembly.

sto-rr~age r~ack- in UI I-4t- 2 and'- 3. Regio n 3 is co .v-mprise of,- fou"r,,,9x3 sto"rage,- and oneh 9v9 stor,-age,,-

r,.ack. in- Unit- 1. 4Refue~l asse. mblies- may be kstoredw in evry,,n, Regio"-n 3 ,.-c ll loation4,'

no cell blocking devices are installed in Region 3.Regions 2 andi A are mived- rand4 are seve\-n 9v8 ra-,ck and three 12x r~acks,,

in Un-it -2c and 3,Regio~ns 2 and I nar m'ixed and are of sixv 9v8 storagare mixed, in a repatnga-n 3vA storwage attern in w.hic-h twor outof cell locations are designated Region-,,

2 and ten outof twelve! cell. locations

..... designated Region I (see-, UFS^RFigres 1 "7 and 9.1 7A). S1nc ..-fuel asse-,mblies may, bestored in e,.ery Region 2 and Region '1 cell location no.. celblocking de,,ice ..... installed in Region 2 =an Region Isecontinued PALO VERDE UNITS 1,2,3B3.171RVSO 4B 3.7.17-1REVISION Insert I for TS Bases 3.7.17 page B 3.7.17-1The spent fuel storage cells are installed in parallel rows with a nominal center-to-center spacing of 9.5 inches. This spacing, a minimum soluble boron concentration of 1460 ppm, theuse of neutron-absorbing panels, and the storage of fuel in the appropriate region based on fuelassembly initial enrichment, discharge burnup, and decay time in accordance with TS Tables3.7.17-1 through 3.7.17-5 is sufficient to maintain keff< 0.95 for fuel of initial maximum radiallyaveraged enrichment of up to 4.65 wt%.Disused CEAs, in-core instruments, and other material is stored in trash cans. A trash can maybe stored in any location that is approved to store a fuel assembly.

No special nuclear material(SNM) may be stored in a trash can.

IAfter SFP transitionI Spent Fuel Assembly StorageB 3.7.17BASESBAAC"KG RUDf/ KF(conti4nued)4 The, fucl storage cclls installed4 in parallel rows,wit4h a nominal cntne-r to nnl cnter-spacing of 9. innHch. ThLspacingn a

slubklc, boron co'-nceant-ration of ppmassembkl,,

b 'urnp i/n accordance4

... th. 4 -.T--

3.17 1,_3.7.17 2, andl 3.T.1T 3 is sufficie'-4nt-to maintain keTT-of0.9 foK r, fueJl ofz orgna,, nl maxi{mum radia,4:lly.

ofi upn to- /1.0/.APPLICABLE Thc ...nt fu-el strge....

pol is dest"igned,4 for' nonrlSAFETY ANALYSES critcalty' by ' "se of

spacing, credi.t-for. boron and3'1 tlhe "fueali4n theappn,'nropriate-TS: FZ{TiguresC 3.7.17 1, 3.7.17 2 3,4 an 37117 3. Th& -.de-l-gmke, i 1.0 assu-mi ng no credi t for boron and _4 Q--_c_taki"ng credit- fo sonclulehl boron. The burnup The analysis of-effe'tII'ct of fuel_ storage9 inInsrt 1th fue,,l soag k-vr acks whas pefome by BB :Combust/ion Engineering.-

(CE)usin the threen,44n dime.'-,.,-nsinlMot-arocd hEO A with th-e updated #an-1f~

ENDF/ neutron:-l cros'/Insert section.

library.....Th E.....code ha been....

previously..used.by L I~~~C for the, analysis of~n fuel 4-rack' rea~ctivity~ haver ben.,,-,.,,

bencmark\I d agaih l-ns resu:3lts, fAn-,rom numerousR c: nritical storagem, ra:lck as as possblel w.,ith respectn4 toThe of DRegio~ns

2. :3,nd4 '1 inc'ludeda

seeal, neglecPted4 theTheeassumptoions',n

'tend4 to increnase tlhecalculated3-n, e,-ffective,,,

multlication-:34,-,,

facto',r-,

(keT of t',-heSThe stored fuel assembli!es w.ere modeled as CE 1..x16vassembl'ies writh a3 nomilnal p'itc-h of 0.5065 "inches betweean fuealrods,,4 a fuerl pellet-of 0.325 inhe,,",ac nd a3 llO2)density of 10.31 g/cc.(conInIued) lbUPALO VERDE UNITS 1,2,3B371-2RVSO B 3.7.17-2REVISION

%

Insert 2 for TS Bases 3.7.17 page B 3.7.17-2The nuclear criticality safety analysis in References 1 and 2 considered the following reactivity-increasing accidents:

• Misload of a single assembly into an unacceptable storage location* Multiple assemblies misloaded in series due to a common cause* Spent fuel pooi temperature outside the allowable operating range* Dropped and misplaced fresh fuel assembly* Seismic event* Inadvertent removal of a NETCO-SNAP-IN~

rack insertIn each case, the spent fuel assembly storage met the requirements of 10 CFR 50.68(b)(4).

Thus, the spent fuel storage facility is designed for noncriticality by use of adequate spacing,and neutron absorbing panels considering initial enrichment, fuel burnup, and decay time.

IAfter SEP transitionI Spent Fuel Assembly StorageB 3.7.17BASE[SAPPLICTfABLEI

,AI'E'TV

,ANAILVSES versus. initial1 enrichmen fo dccay tims in 5 ..e.rincrements from 0 to 20 years for both Regions 3 and 4(TS 3.7.17 2 and 3.7.17 3) uciih that- all poimnt-s ornt'he curves a kr valu(inc/mludingr all bi.ases .andu-ncertainties) of 1.0 for.. unboratcd wae...assccriated,-

wit4h met-hodo"log nd wm- .ater, tenmpr~n3-atre wereinluedIcl...

and unerante assuciaedtainthiethodologyat QKEO nr',a calculation fuel, enrichme, nt,l fue rckl pitch, fuel.-,

requret maintainthespenful pool1 ....he1,, (Incui ngm'-4m' allC biasesP un'ertaF'inties) 4h~

95 r, obabiCll ty/95,%PIP confidene-h.-

leve.3P ,--A,,4!

soluble, boromn Hco4ncentratmiokn of,900ppm i eurdt sueta h pnMost' abnr,,,,mal

storage, will not-H resuclt in anincreasea in the ke, of te racks.~ Howevr.,

, it- is possibletlhe aceptab-3le area in TS 3.7.17 1.',, orwitht-he accepntable area in TS Figuresc 3.7.17 2 or 3.7.17 3, w.hichcould latonincrease inreactivity.

These evnt wulincluden an asse.,,mbly.

dropr-topof a rac orI a rack,the pool walls, or the misloading of a~n assembly.

Foboron.in the spent fuel poo water to ensure..protection requ~l'ire tlhe of'- tl-wo' unlikenly, indepndnt,~mn solubl boron concentratio of only 900 ppm is require tomassure. that h, remain 4o 0.9 assuming.

t£. he single. mostfuel poom'l water.P As schl," t-he reductiofnm in h,, Pcauseby%1 required4 sof' lub~le boroPn concentration'{'

t'han staff criterion of ke, _ 0.9 is met at all times.PALO VERDE UNITS 1,2,3B37.73RVSO B 3.7.17-3REVISION

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IAfter SFP transitionI Spent Fuel Assembly StorageB 3.7.17BASESADDI The criticlity a.pect. of t he spent fuel pool meet theThe spent poo,, heat load, calculations w;ere based on afu, ll pool 1205F fuenl assembl-ies.

Frm t-rhe spent fueilpool0+',

criticalt nalsiThem Mnumber~-

of fueln assemblies, thatl can beC in~ thel~ four- reginl =configuatin i1209Efue asse--imblies.

basi oMf thel spntr fuelare mehnhmica=lly blockedH to -imait the mav-imum number oxf fuenlassemblies t-hat- mayl be3 sltoreda

-in the spntl fueil poollomading of the pool. The fuel pool rac~k coinsrctmlion than. oine assmbl fromA being, impacted uionaful andrling ah .c¥cieant, Th UFSAR analysnis 3cncnml u -ion .hregardn g worst-f scnai for, a dropped assembly (il-n whch-,nifuenl assembly, damages fuel rodsc in the dropped~,-

assembly, not -imnrImr"'

fuel "in tlhe st-ored assemhliesncontinues,n toThe spent fuel assembly storage satisfies Criterion 2 of 10CFR 50.36 (c)(2)(ii).

I abes3..1-Itheog spent7- fulpoa awyeand Fi1u0e3assuming the.LCOare rsictonsitn woh the crtcli~~ety safetye anlslies performefrthe spent f el pool, according to Figure 3.7.....

1.....and 3..17 3in in theaccmpanying LCO,.nue ht h efoSpecification 4.3.1.1 provides additional details for fuelstorage in each of the Regios(conti nued)PALO VERDE UNITS 1,2,3 B 3.7.17-4 REVISION IAfter SFP transition Spent Fuel Assembly StorageB 3.7.17BASESAPPLICABILITY This LCO applies whenever any fuel assembly is stored in thespent fuel pooi.ACTIONSA.1I Tables 3.7.17-1 through 3.7.17-5 and Figure 3.7.17-1'-7I v wRequired Action A.1 is modified by a Note indicating tLCO 3.0.3 does not apply. /When the configuration of fuel assemblies s ed the spentfuel pool is not in accordance with Fgrs 3.7.17 2, and 3.7.17 3, immediate action must; taken tomake the necessary fuel assembly movement(s) bring theconfiguration into compliance with N7!3.7.17-2, and 3.7.17 3.If moving irradiated fuel assemblies while in MODE 5 or 6,LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1. 2. 3. or 4, the fuelmovement is independent of reactor operation.

Therefore, ineither case, inability to move fuel assemblies is notsufficient reason to require a reactor shutdown.

SURVEILLANCE REQU IREMENTSThis SR administrative means that the initialenrichment of the fuel assembly is in accordance with 3 71!7 1 3_717 2. and 3_7.7 3 in theofTbe371-thurough 3..1-5 performanc ofh thi SR.....

-ensure compliance with Specification 4.3.1.1.

l(ot udPALO VERDE UNITS 1,2,3B371-5RVSO B 3.7.17-5REVISION

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IAfter SFP transitionI Spent Fuel Assembly StorageB 3.7.17BASES requ~lirement-fo~r 2 must e'-h sltored a-in Regnion 1,(continRued)

• that anyJ fuel assembly/

may/ be stored a-in aRegion 1,4-tht- any\ "fuel assembly, may be stored -in a loer, nuimbereda regionn t-han the rego w: .h-ic-h it- qalifies bc au-se~~~~REFERENE

.USR aSo oTc peci0n 9 1.3 n .1.3),* n thatI_ act al bur ,nup '- to the burnup -,A+,"reurmn o eodcytiewllasb orc4.- 13g_ N 001190 1221 1, "Palo,,"

\/erdeH Spent- Fuel Pool/Cr'4-it,-1icality Analysi,c

" ,ABBclulto PV FE/ 0106,Evaluatio-4n ofc the Spntn4 Fuel Pool Map w.,it-h a Proposedca Reg-,ionr 3 Inrease, ac" datedH Febrary, 25-, 2010.-"Criticality Safety Analysis for Palo Verde Nuclear Generating Station Units 1, 2, and 3" (Proprietary),

WCAP-1 8030-P,Revision 0, September 2015.PALO VERDE UNITS 1,2,3 B 3.7.17-6 REVISION Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 4List of Regulatory Commitments Enclosure Description and Assessment of Proposed License Amendment Attachment 4List of Regulatory Commitments Reuaor omimnDue Date/Milestone 1. APS will implement procedural controls torequire verification that fresh fuelassemblies are not placed face-adjacent to one another before completing a fuelmove.2. The transition to the new SEPconfiguration will be completed in all threeunits in accordance with the Spent FuelPool Transition Plan within two years ofthe NRC approval date of the amendment or by December 31, 2019, whichever islater.Within the requested 90-day implementation period following NRC approval.

Within two years of the NRC approval date ofthe amendment or by December 31, 2019,whichever is later.