ML19275E310

Submittal of License Amendment Request to Update the Spent Fuel Pool Criticality Safety Analysis and Revise Technical Specification (TS) 3.7.15 Spent Fuel Assembly Storage and TS 4.3 Fuel Storage

ML19275E310
Person / Time
Site:	Farley
Issue date:	09/30/2019
From:	Gayheart C Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML19275E393	List: ML19275E310 NL-19-0796, WCAP-18414-NP, Rev 0, J.M. Farley Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis (Non-Proprietary Version)
References
NL-19-0796
Download: ML19275E310 (57)

Text

A Southern Nuclear September 30, 2019 Docket Nos.: 50-348 50-364 ATTN : Document Control Desk U. S. Nuclear Regulatory Commission Washington, D. C. 20555-0001 Cheryl A. Gayheart Regulatory Affairs Director Joseph M. Farley Nuclear Plant Units 1 and 2 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5316 tel 205 992 7601 fax cagayhea@southernco.com 10 CFR 50.90 NL-19-0796 Submittal of License Amendment Request (LAR) to Update the Spent Fuel Pool Criticality Safety Analysis and Revise Technical Specification (TS) 3.7.15 "Spent Fuel Assembly Storage" and TS 4.3 "Fuel Storage" Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) requests an amendment to the Joseph M. Farley Nuclear Plant (FNP) lJnit 1 Renewed Facility Operating License (NPF-2), and Unit 2 Renewed Facility Operating License (NPF-8), by incorporating the attached proposed change into the Unit 1 and Unit 2 Technical Specifications (TSs).

Specifically, the proposed change is a request to revise TS 3.7.15 and TS 4.3 to:

allow for an updated spent fuel pool (SFP) criticality safety analysis; and account for the impact on the spent fuel from a proposed measurement uncertainty recapture (MUR) power uprate.

The Enclosure provides a description and assessment of the proposed changes. Attachment 1 provides the existing TS pages marked to show the proposed changes. Attachment 2 provides retyped TS pages. Attachment 3 provides existing TS Bases pages marked to show the proposed changes for information only. Attachment 4 provides a proprietary version of the technical evaluation [WCAP]. Attachment 5 provides a non-proprietary version of the technical evaluation [WCAP]. Attachment 6 provides the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-19-4943, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice.

Approval of the proposed amendments is requested by October 1, 2020 to support the Fall 2020 refueling outage for FNP Unit 2. The proposed changes will be implemented within 90 days of issuance of the amendment.

The Fall 2020 refueling outage is the implementation date for the FNP MUR power uprate modifications. SNC plans to submit the MUR LAR by the end of October 2019.

SNC to NRC SFP Criticality LAA NL-19-0796 Page 2 This letter contains no NRC commitments.

In accordance with 10 CFR 50.91, SNC Is notifying the state of Alabama of this license amendment request by transmitting a copy of tflis letter to the designated state official.

If you have any questions, please contact Jamie Coleman at 205.992.6611.

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

Executed on September 30, 2019.

C.A a Regul t flairs Director Southern Nuclear Operating Company efb/scm

Enclosure:

Description and Assessment of the Proposed Changes : Technical Specification Page Markups : Retyped Technical Speclflcatlon Pages : Technical Specification Bases Page Markups - For Information only : WCAP-18414-P "J.M. Farley Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis* (Proprietary Version) : WCAP-18414-NP "J.M. Farley Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis" (Non-proprietary Version) : Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-19-4943, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice cc:

NRC Regional Administrator NRC NRR Project Manager - Farley 1 &2 NRC Senior Resident Inspector-Farley 1 & 2 Alabama - State Health Officer for the Department of Public Health SNC Document Control R-Type: CFA04.054

SNC to NRC LAR Enclosure NL-19-0796 ENCLOSURE Description and Assessment of the Proposed Changes

Subject:

Joseph M. Farley Nuclear Plant Units 1 and 2 Submittal of License Amendment Request to Revise Technical Specification (TS) 3.7.15 "Spent Fuel Assembly $torage" and TS 4.3 "Fuel Storage"

1.

SUMMARY

DESCRIPTION

2.

DETAILED DESCRIPTION 2.1

'System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change 2.4 Description_ of the Proposed Change

3.

TECHNICAL EVALUATION

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination Analysis 4.4 Conclusions 1

5.

ENVIRONMENTAL CONSIDERATION

6.

REFERENCES ATTACHMENTS

1.

Technical Specification Page Markups

2.

Retyped Technical Specification Pages

3.

Technical Specification Bases Page Markups - For Information Only

4.

WCAP-18414-P "J. M. Farley Units 1 and 2 Spent Fuel Pool Criticality Safety ~nalysis" (Proprietary Version),

5.

WCAP-18414-NP "J.M. Farley Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis" (Non-proprietary Version)

6.

Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-19-4943, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice E-1

SNC to NR._C LAR Enclosure NL-19-0796

1.

SUMMARY

DESCRIPTION Southern Nuclear Operating Company (SNC) requests an amendment to the Joseph M. Farley Nuclear Plant (FNP) Unit 1 Renewed Facility Operating License (NPF-2) and Unit 2 Renewed Facility Operating License (NPF-8) by incorporating the attached spent fuel storage changes into the FNP Unit 1 and Unit 2 Technical Specifications. The license amendment request proposes changes to spent fuel storage Technical Specification (TS) 3. 7.15 "Spent Fuel Assembly Sto,rage" and TS 4.3 "Fuel storage" for FNP Unit 1 and Unit 2. The purpose of the proposed changes is to allow for an updated spent fuel pool criticality safety analysis and to account for the impact on the spent fuel from a proposed measurement uncertainty recapture (MUR) power uprate.

2. DETAILED DESCRIPTION 2.1 System Design and Operation The spent fuel pool (SFP) is made up of one fuel storage rack design (region) that maintains

10. 75-inch ~nter to center spacing between spent fuel assemblies. The Farley Units 1 & 2 SFPs each consist of two 6 x 7, nineteen 7 x 7, and seven 7 x 8 storage racks. The spent fuel racks are freestanding and are free to move on the pool liner floor during a seismic event.

The SFP storage capacity is 1,407 fuel assemblies. The actual storage capacity is limited by Technical Specification 3.7.15, Spent Fuel Ass'embly storage, and is dependent upon the fuel characteristics of the FNP fuel inventory.

The storage racks -ve of flux trap style with an uncredited Boraflex neutron absorber panel on the sides of each storage cell., This results in a flux trap between any two assembly storage locations. No credit is taken for the.presence of residual Boraflex in the current licensing basis per-Section 4.3.2. 7.2.1 and 4.3.2 7.2.2 of the FNP Units 1 and 2 Updated Final Safety Analysis Report (UFSAR). The burnable absorber cavity is assumed to be filled with water of the same composition as the water elsewhere in the storage racks.

Per Section 9.1.2.1 of the FNP Units 1 &2 UFSAR, the spent fuel racks are designed to withstand shipping, handling, and normal operating loads (impact and dead loads of fuel assemblies) as well as safe shutdown earthquake (SSE) and one-half SSE seismic loads meeting American Nuclear Society (ANS)"Safety Class 3 and American Institute of steel Construction (AISC) requirements. The spent fuel racks are also designed to meet the Category 1 seismic requirements of Regulatory Guide 1.13.

The updated criticality safety analysis,

• J. M. Farley Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis" (Attachment 4 - WCAP-18414-P, Proprietary Version and Attachment 5 -

WCAP-18414-NP, Non-Proprietary Version) evaluates the SFP storage racks for the placement of fuel within the storage arrays defined in the technical specifications. Credit is taken for the negative reactivity associated with bumup and post-irradiation cooling time (decay time) for assemblies which have been operated in the reactor. Fuel assemblies which have riot operated in the reactoF-may take credit for the presence of zirconium diboride in the integral fuel burnable absorber (IFBA). While the FNP Unit 1 and Unit 2 SFP storage racks may contain Boraflex absorber inserts, no credit is taken for the presence of Boraflex absorber as described in the UFSAR However, credit is taken for the presence of soluble boron in the SFP.

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SNC to NRC LAR Enclosure NL-19-0796 2.2 Current Technical Specifications Requirements SNC proposes changes to the spent fuel storage and affects Technical Specification (TS) 3. 7.15 "Spent Fuel Assembly Storage" and TS 4.3 "Fuel Storage" for FNP Unit 1 and Unit 2.

2.3 R~ason for the Proposed Change The purpose of the proposed changes to TS 3. 7.15 and TS 4.3 is to update the SFP criticality safety analysis. The updated analysis, per the NRC's request, follows the guidance in NEI 12-16 (Reference 6.1). The proposed changes also account for the impact on the spent fuel from a proposed measurement uncertainty recapture (MUR) power uprate.

2.4 Description of the Proposed Change TS 3.7.15, "Spent Fuel Assembly Storage" The proposed change revises TS 3.7.15 based on the updated criticality safety analysis, "J. M.

Farley Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis" (Attachment 4-WCAP-18414-P, Proprietary Version and Attachment 5 - WCAP-18414-NP, Non-Proprietary Version) and the proposed changes to TS 4.3.

The proposed changes to TS 3. i 15 include:

removal of the reference to "within the Acceptable Bumup Domain of Figure 3.7.15-1",

removal of the reference to "Figure 3.7.15-1 or" in Surveillance Requirement (SR) 3.7.15.1, and 1 removal of TS Figure 3.7.15-1 which contains the fuel assembly bumup limit requirements for all cell storage. This figure will be replaced by Figure 4.3-1, Table 4.3-1, and Tables 4.3-3 through 4.3-5 in Section 4.3.

Note: Attachment 4 and Attachment 5 are applicable to all references to the updated criticality safety analysis from hereon.

TS 4.3, "Fuel Storage" The proposed change to TS 4.3 revises the section based on the updated criticality safety analysis. The proposed changes include updated fuel storage configurations (arrays) and restrictions for storage of fuel in the storage racks based on fuel assembly bumup limit requirements.

The proposed changes to TS 4.3 include:

Replacement of Specification 4.3.1.1.e with "New or partially spent fuel assemblies that must be stored according to their combination of discharge bumup and nominal enrichment, decay time since operation, required IFBA (if applicable), a'nd must comply with Figure 4.3-1, Table 4.3-1, and Tables 4.3-3 through 4.3-5 (as applicable). Each assembly should be stored in an appropriate storage configuration according to its fuel category as specifically described in Table 4.3-1 and geometry based on Figure 4.3-1 ;"

Replacement of Specification 4.3.1.1.f with MFuel assemblies that are stored in accordance with every applicable storage array as shown in Figure 4.3-1 of which they I

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SNC to NRC LAR Enclosure NL-19-0796 are a part (i.e., one fuel assembly cah be part of up to four different storage arrays, each storage array shall be in accordance with Figure 4.3-1); and" Replacement of "Figure 4.3-6" with "Figure 4.3-2" in Specification 4.3.1.1.g.

Replacement of Figures 4.3-1.'through 4.3-5 with proposed Figure 4.3-1 and Tables 4.3-1 through 4.3-5.

o The proposed Figure 4.3-1 contains graphical and verbal descriptions of the four allowed storage arrays.

o The proposed Table 4.3-1 provides interpretation of the fuel categories included in Figure 4.3:-1 and it includes references to the appropriate table that provides the fitting coefficients that are to be used to calculate the minimum required fuel assembly bumup or the minimum IFBA requirement.

o The proposed Table 4.3-2 provides the maximum enrichment allowed for each fuel category with 0.0 MWd/MTU bumup.

o The proposed Table 4.3-3 provides the fitting coefficients to calculate the minimum required fuel assembly bumup for fuel categories 3 and 4 for Standard Fuel Assembly (STD)/Robust Fuel Assembly (RFA) fuel.

o The proposed Table 4 3-4 provides the fitting coefficients to calculate the minimum required fuel assembly bumup for fuel categories 3 and 4 for Optimized Fuel Assembly (OFA) fuel.

o The proposed Table 4.3-5 provides the fitting coefficients to calculate the minimum IFBA requirements for fuel category 2.

Renumbering Figure 4.3-6 as Figure 4.3-2. The specifications that reference Figure 4.3-6 are being changed to reference Figure 4.3-2.

Redline/strikeout copies of TS 3. 7.15 and TS 4.3 are inctuded in Attachment 1. Retyped copies TS 3.7.15 and TS 4.3 are provided in Attachment 2. Redline/strikeout copies of TS 3.7.14 Bases and TS 3.7.15 Bases are provided (for information only) in Attachment 3.

3. TECHNICAL EVALUATION This License Amendment Request, at the NRC's request, was modeled based on the guidance of NEI 12-16 (Reference 6.1). SNC used the NEI guidance to ensure that the proper considerations were made in the analysis and controls. The updated criticality safety analysis also reflects the guidance of DSS-ISG-2010-001, "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pool" (Reference 6.2).

This section includes a brief statement related to each applicable topic discussed in NEI 12-16 (Reference 6.1) and summarizes the analysis or proposed controls applicable to each area.

Also included is a discussion on the proposed Technical Specification changes. Supporting details are found in Attachments 4/5, as referenced below.

Acceptance Criteria NEI 12-16, Section 2 (Reference 6.1) describes the NRC acceptance criteria for spent fuel pool storage of new and used fuel for pools where credit for soluble boron is taken as follows:

1. The criticality safety analyses must meet two independent limits:

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SNC to NRC LAR Enclosure NL-19-0796

a. Wrth the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water, the ke11 must remain below 1.0 (subcrttical), at a 95 percent probability, 95 percent confidence level.

b. Wrth the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity and flooded with borated water, the i<eff must not exceed 0.95, at a 95 percent probability, 95 percent confidence level.

Section 2.1 of the updated criticality safety analysis in Attachment 4 provides the same acceptance criteria for the updated FNP SFP criticality analysis and carries forward the criteria approved in the NRC Safety Evaluation issued with the FNP SFP criticality analysis WCAP-14416-P approved in Amendment No. 133 to Facility Operating License No. NPF-2 and Amendment No. 125 to Facility Operating License No. NPF-8 for Joseph M. Farley Nuclear Plant, Units 1 and 2 (Accession No. ML013130226).

Computer Codes NEI 12-16, Section 3 (Reference 6.1) describes the different types of computer codes that may be used in a criticality analysis. This section also discusses the validation of the computer codes used in the criticality analysis. The licensee needs to state which codes were utilized along with the type/version of cross-section libraries.

The updated criticality safety analysis utilized in this submittal employs the following computer codes and cross-section libraries:

(1) The twcx1imensional (2-D) transport lattice code PARAGON Version 1.2.0 and its cross-section library based on Evaluated Nuclear Data File Version Vl.3 (ENDF/8-Vl.3).

(2) Scale Version 6.2.3 with the ENDF/B-Vll 238-group cross-section library.

The Computer Codes used in this application are discussed in Section 2.3 of the updated criticality safety analysis. Paragon is generically approved for depletion calculations. The applicable specific validation of Scale Version 6.2.3 is provided in Appendix A of the updated criticality safety analysis.

Reactivity Effects of Depletion NEI 12-16, Section 4 (Reference 6.1) describes appropriate considerations for calculating reactivity effects of fuel depletion. Significant parameters that could impact reactivity of used fuel in depletion analyses are:

Power, moderator temperature and fuel temperature during depletion Soluble boron during depletion Presence of burnable absorbers Rodded operation Cooling time and other depletion parameters are also discussed in NEI 12-16, Section 4 (Reference 6.1).

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$NC to NRC LAR Enclosure NL-19-0796 The depletion analysis is described in Section 4 of the updated criticality safety analysis and describes the methods u$ed to determine conservative and bounding inputs for the generation of isotopic number densities. Controls which ensure Mure fuel designs satisfy the assumptions of the analysis are discussed in the Licensee Controls section, below. Depletion uncertainty 1s discussed in Section 5.2.3.1.5 of the updated cnticality safety analysis.

Fuel Assembly and Storage Rack Modeling NEI 12-16, Section 5 (Reference 6.1) describes generally acceptable methods of modeling fuel assemblies and fuel storage racks, including considerations for rack neutron absorbers.

Two fuel assembly designs are used at Farley and are considered in the updated criticality safety analysis, the Westinghouse standard fuel assembly (STD) and the Westinghouse optimized fuel assembly (OFA). The Westinghouse robust fyel assembly (RFA) is also considered in the updated criticality safety analysis to support potential future use. The burnup requirements developed for the STD fuel design are applicable to the RFA fuel design because the neutronically important characteristics are the same. Details of the fuel assembly designs are provided in Section 3.1 and 4.3.1 of the updated criticality safety analysis. The updated criticality safety analysis allows fuel assemblies which have not operated in the reactor to take credit for the presence of zirconium diboride in the integral fuel burnable absorber (IFBA).

The spent fuel pool is made up of one fuel storage rack design (region). The Farley Units 1 & 2 SFPs each consist of two 6 x 7, nineteen 7 x 7, and seven 7 x 8 storage racks. The storage

\\ racks are of flux trap style with an uncredited Boraflex neutron absorber panel on every side of each storage cell. This results in a flux trap between any two assembly storage locations. No credit is taken for the presence of residual Boraflex. The burnable absorber cavity is assumed to be filled with water of the same composition as the water elsewhere in the storage racks. Credit is taken for the presence of soluble boron in the SFP. Details of the storage rack parameters are provided in Section 3.2 of the updated criticality safety analysis.

The fuel and storage rack manufacturing tolerances, eccentric fuel assembly positioning bias,

• and SFP temperature bias are included in the updated criticality safety analysis through either analysis or use of bounding values. Details of these items are provided in Section 5.2.3.1 of the updated criticality safety analysis The axial bumup distribution and reactor record bumup uncertainties are considered in the

\\

updated criticality safety analysis. Details of the axial bumup distribution are provided in Section 4.2.3. The details for the bumup uncertainties are provided in Section 5.2.3.1.4.

The new fuel storage racks were not considered in the updated SFP criticality safety analysis update. This license amendment involves no changes to the design or operation of the new fuel storage racks.

Configuration Modeling and Soluble Boron Credit NEI 12-16, Section 6 (Reference 6.1) describes considerations for configuration modeling, including a description of nonnal conditions, interface considerations, and abnonnal / accident conditions which should be considered. NEI 12-16, Section 7 (Reference 6.1) describes considerations for soluble boron credit under nonnal and accident conditions, and considerations for a boron dilution accident.

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SNC to NRG LAR Enclosure NL-19-0796 V

The updated criticality safety analysis demonstrates acceptable results for keff for both normal conditions (Section 5.4) and accident conditions (Section 5.5.2). Normal conditions inciud_e normal storage, fuel movement, and other procedurally controlled activities in the SFPs. Fuel assembly storage arrays which define allowable storage are defined in Section 5.2.1 of the updated criticality safety analysis, which have been incorporated into the proposed Technical Specification changes. Controls which ensure that the proposed Technical Specification limitations for storage are maintained are discussed in the Licensee Controls section below.

Interface considerations are described in Section 5.3 of the updated criticality safety analysis.

The interfaces are the locations where there is a change in either the storage racks or the storage requirements of the fuel in question. Only the intra-region interfaces are evaluated because all racks are of the same design and no pool region interfaces are present. Each storage cell can be part of four different storage arrays. Compliance with the storage arrays in the technical specifications will ensure acceptable boundary cells at the interface.

Per Section 4.3.2. 7.2.1 of the PNP Units 1 &2 UFSAR, most postulated accidents in the spent-fuel rack will not result in an increase in reactivity. These include dropping an assembly on top of the rack (rack structure maintains 10 in. of separation befyieen dropped and stored assemblies, precluding interaction), or dropping an assembly into a position other than a storage cell (prevented by design of rack). However, accidents can be postulated for each storage configuration which would increase reactivity beyond the analyzed condition. The first postulated accident would be a loss of the fuel pool cooling system. The second accident would be dropping an assembly into an already loaded cell and the third would be a mislead of an assembly into a cell for which the resbictions-on location, enrichment, or bumup are not satisfied.

(

Single mislead events were previously analyzed; however, the updated criticality safety analysis also includes analysis of a multiple mislead accident scenario in accordance with NEI 12-16, Section 6.3.5. The inclusion of this analysis does not imply the creation of the possibility of a new accident, but expands the boundaries of the analyzed accident conditions to ensure that all potential accidents are properly considered.

For the limiting normal condition, 320 ppm of soluble boron is credited to ensure the maximum ketf satisfies the acceptance criteria of~ s 0.95.

The limiting analyzed accident condition was an event which involves misleading multiple fuel assemblies in series due to a common cause.

In the multiple mislead accident cases, 1710 ppm of boron is required to maintain ketr s 0.95.

This amount of boron is bounded by the current limit in Technical Specification 3.7.14 "Fuel Storage Pool Boron Concentration" which requires greater than 2000 ppm of boron concentration in the fuel storage pooi'.

A spent-fuel pool dilution event has been previously evaluated by SNC and determined to not be a credible event for Farley Units 1 & 2.

• Section 4.3.2.7.2.1 of the FNP Units 1&2 UFSAR,

provides the following summary:

"A spent-fuel pool boron dilution evaluation was performed to determine the volume necessary to dilute the spent-fuel pool from the Technical Specification limit of 2000 ppm to 400 ppm (the boron concentration required to maintain l<etf s 0.95). The boron dilution evaluation determined that approximately 480,000 gal of water would be required to dilute E-7

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SNC to NRC LAR Enclosure NL-19-0796 the spent-fuel pool from 2000 ppm to 400 ppm. A dilution event that would result in this large volume of water would require the transfer of a large quantity of water from the dilution source and a significant increase in the spent-fuel pool level, which would ultimately overflow the pool.

This large volume of water would be readily detected and terminated by plant personnel. A spent-fuel pool dilution event of this magnitude is not a credible event."

SNC's previous evaluation was considered and accepted in the NRC Safety Evaluation for Amendments 133/125, and therefore, the spent-fuel pool dilution event was not re-analyzed as part of the updated criticality safety analysis.

The abnormal/accident section of NEI 12-16, Section 6 (Reference 6.1) also includes seismic events. Section 5.5.2.4 of the updated criticality safety analysis addresses the seismic event. In the event of an earthquake or similar seismic event, the SFP storage racks can shift position.

This can cause the rack modules to slide together eliminating the space between modules and between modules and the spent fuel pool wall. The fuel assembly position analysis In the updated criticality safety analysis covers possible water gap reduction between assemblies due to a seismic event. Similar to the multiple mislead accident scenario, the inclusion of this analysis did not imply the creation of the possibility of a new accident but expands the -

boundaries of the analyzed accident conditions to ensure that all potential accidents are properly considered. The effects of this event are bounded by the worst-case fuel assembly misleading event.

Calculation of Maximum Kett NEI 12-16, Section 8 (Reference 6.1) describes that the maximum keff is determined by adding to the nominal calculated ketr any biases that may exist in the methodology and the applicable uncertainties using the formula described below, for comparison to the acceptance limits.

m n

km.ax = keti + L Bias1 + L Uncertainty/

i=O J=O The updated criticality safety analysis demonstrates that the !<err, including all applicable biases and uncertainties which account for the statistical 95/95 confidence levels, satisfy the acceptance criteria. The sum <;>f biases are additive while the sum of uncertainties are statistically added as the root sum square of the individual reactivity uncertainties as described in Sections 5.2.2 and 5.2.3 of the updated criticality safety analysis.

Licensee Controls NEI 12-16, Section 9 (Reference 6.1) describes controls Intended to ensure that the conditions evaluated in the nuclear criticality safety analysis are and remain bounding to the current plant operating parameters. It discusses procedural controls for fuel storage and for planning and performance of fuel movements, new (future) fuel types, and pre-and post-Irradiation fuel characterization.

In conjunction with the implementation of the approved LAR, the controls are changed to be consistent with the updated criticality safety analysis. These controls ensure the spent fuel pool E-8

SNC to NRG LAR Enclosure NL-19-0796 configuration and other applicable conditions evaluated in the updated criticality safety analysis remain bounding when compared to current fuel design and plant operating parameters.

Specifically, these controls ensure:

t. TS 3.7.15 and TS 4.3 compliance is maintained at all times. Controls are established to ensure that all fuel movement plans into the spent fuel pool are prepared in a manner which ensures continual compliance with the limitations of proposed TS 3.7.15 and TS 4.3, including all intermediate steps during fuel movement.

2. A misleading event beyond the analyzed accident conditions is not credible. Controls are established to ensure that an error in the fuel move planning does not have the potential to result in a misleading accident which is not bounded by the updated criticality safety analysis.

3. Assumptions related to fuel characterization and reactor operation remain valid. Controls are established to ensure that conditions evaluated in the updated criticality safety analysis will remain bounding for both Mure fuel design changes (pre-irradiation fuel characterization) and future operating conditions (post-irradiation fuel characterization).

Conclusion In conclusion, the proposed changes to TS 3.7.15 and TS 4.3 allow for continued safe storage of spent fuel at Farley Units 1 and 2.

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The Commission's regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36. The TS requirements in 1 O CFR 50.36 include the following categories: (1) safety limits, limiting safety system settings, and limiting control settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.

The requirements for system operability during movement of irradiated fuel are included in the TSs in accordance with 10 CFR 50.36(c)(2), Limiting Conditions for Operation. As required by 10 CFR 50.36(c)(4), design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1),

(2), and (3) of 10 CFR 50.36. This amendment request concerns 10 CFR 50.36(c)(2) and 10 CFR 50.36(c)(4).

At a regulatory level, 10 CFR 50.68(a) requires licensees to select one of two options to satisfy criticality accident requirements: (1) 10 CFR 70.24, or (2) 10 CFR 50.68(b) as highlighted in RIS 2005-05, "NRG Regulatory Issue Summary 2005-05 Regulatory Issues Regarding Criticality Analyses for Spent Fuel Pools and Independent Spent Fuel Storage Installations," dated March 23, 2005.

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SNC to NRG LAR Enclosure NL-19-0796 SNC was granted exemption from 1 O CFR 70.24 in NRG letter titled "RequestJor Exemption from 1 O CFR 70.24 Criticality Monitoring Requirements - Joseph M. Farley Nuclear Plant, Units 1 and 2 (TAC NOS. M95496 and M95497)", dated July 31, 1996.

The critical accident requirements for the current FNP SFP criticality analysis WCAP-14416-P were approved by the NRG in Amendment No 133 to Facility Operating Ucense No. NPF-2 and Amendment No. 125 to Facility Operating License No. NPF-8 for Joseph M. Farley Nuclear Plant, Units 1 and 2 (Accession No. ML013130226). These requirements are also used in the updated Criticality Safety Analysis (Attachment 4/5).

As guidance for reviewing criticality analyses of fuel storage at light-water reactor power plants, the NRC staff issued an internal memorandum on August 19, 1998 (ADAMS Accession No. ML003728001). This memorandum is known as the "Kopp Letter." The Kopp Letter provides guidance on s'alient aspects of a criticality analysis. The guidance is germane to boiling-water reactors and pressurized water reactors, and to borated and unborated conaitions.

On September 29, 2011, the NRC staff issued the Interim Staff Guidance (ISG) DSS-ISG-2010-01, "Staff Guidance Regatding the Nuclear Criticality Safety Analysis for Spent Fuel Pool,"

Accession No. ML110620086. The purpose of the ISG is to provide updated review guidance to the NRC staff to address the increased complexity of recent SFP nuclear criticality analyses and operations. The ISG re-baselines the NRC's expectations for spent fuel criticality analysis. The

, expectations of the ISG were further reinforced in subsequent NRG Information Notice 2011-03, "Nonconservative Criticality Safety Analyses for Fuel Storage," Accession No. ML103090055.

General Design Criterion (GDC) 61 - Fuel storage and handling and radioactMty control, "The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions.

These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions."

• GDC 62 - Prevention of criticality in fuel storage and handling, "Criticality in the fuel storage and handling system shall be prevented by physical systems or processes. Preferably by use of geometrically safe conditions.

Additional guidance is available in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition,"

particularly Section 9. t 1, "Criticality Safety..of Fresh and Spent Fuel Storage and Handling,"

Revision 3, issued March 2007. Section 9.1.1 provides the existing recommendations for performing the review of the nuclear criticality safety analysis of SFPs.

4.2 Precedent

/

The following license amendment requests and applicable RAls have been used in the development of the criticality safety analysis and the appropriate sections of the amendment request:

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SNC to NRC LAR Enclosure NL-19-0796 Comanche Peak Units 1 and 2 License Amendment Request for Spent Fuel Pool Technical Specification Changes in a letter dated July 1, 2014, "Comanche Peak Nuclear Power Plant, Units 1 and 2 - Issuance of Amendments RE: Revision to Technical Specifications 3.7.16, "Duel Storage Pool Boron Concentration," 3.7.17, "Spent Fuel Assembly Storage," 4.3, "Fuel Storage,"

and 5.5, "Programs and Manuals" (TAC NOS. MF1365 and MF1366)," Accession No.'

ML14160A035.

Palo Verde Units 1, 2, and 3 License Amendment Request for Spent Fuel Pool Technical Specification Changes in a letter dated July 28, 2017, "Palo Verde Nucl~ar Generating Station, Units 1, 2, and 3 - Issuance of Amendments to Revise Technical Specifications to Incorporate Updated Criticality Safety Analysis (CAC NOS. MF7138, MF7139, and MF7140)," Accession No. ML17188A412.

Prairie Island Units 1 and 2 License Amendment Request for Spent Fuel Pool Criticality Technical Specification Changes in a letter dated Noverrtber 30, 2017, "Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendment Revising Spent Fuel Pool Criticality Technical Specification (CAC NOS. MF7121 and MF7122, EPID L-2015-LLA-0002)," Accession No. ML17334A178.

4.3 No Significant Hazards Consideration Determination Analysis SNC has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92(c) as discussed below:

1)

• Does the proposed amendment involve a significant increase in the probability or -

consequences of an accident previously evaluated?

Response: No.

The proposed amendment was evaluated for impact on the following criticality events and accidents and no impacts were identified: (1) loss of spent fuel poql cooling system, (2) dropping a fuel assembly into an already loaded storage cell, and (3) the misleading of a single fuel assembly or multiple fuel assemblies into a cell for which the restrictions on location, enrichment, or bumup are not satisfied.

Operation in accordance with the proposed amendment will not change the probability of a loss of spent fuel pool cooling because the changes in the criticality safety analysis have no bearing on the systems, structures, and components involved in initiating such an event. A criticality safety analysis of the limiting fuel loading configuration confirmed that the condition would remain subcritical for a range of normal and accident conditions. The,effects of the accident conditions are bounded by the multiple fuel assembly mislead accident.

• Operation in accordance with the proposed amendment will not change the probability of a fuel assembly being dropped into an already loaded storage cell because fuel movement will continue to be controlled by approved fuel handling procedures. The consequences of a dropped fuel assembly are not changed; there will continue to be significant separation between the dropped fuel assembly and the active regions of the fuel assemblies. The effects of this accident are bounded by the multiple fuel assembly mislead accident.

E-11

)

SNC to NRC LAR Enclosure NL-19-0796 Operation in accordance with the proposed amendment will not change the probability of a fuel assembly misleading because fuel movement will continue to be controlled by approved fuel selection and fuel handling procedures. These procedures continue to require identification of the initial and target locations for each fuel assembly and fuel assembly insert that is moved. The consequences of a fuel misleading event are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for the multiple fuel assembly mislead accident.

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

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

Response: No.

The potential for criticality in the spent fuel pool is not a new or different type of accident.

Storage configurations allowed by Technical Specifications 3. 7.15 and 4.3 have been

• analyzed to demonstrate that the pool remains subcritical.

• The new criticality safety analysis includes analysis of a multiple mislead accident scenario; only single mislead events were previoµsly analyzed. The inclusion of this analysis does not imply the creation of the possibility of a new accident, but simply expands the boundaries of the analyzed accident conditions to ensure that all potential accidents are properly considered.

There is no significant change in plant configuration, equipment design or usage of plant equipment. The updated criticality safety analysis assures that the pool will continue to remain subcritical.

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

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

Response: No.

The proposed change was evaluated for its effect on current margins of safety as they relate to criticality. The margin of safety for subcriticality required by Amendment No. 133 to Facility Operating License No. NPF-2 and Amendment No. 125 to Facility Operating License, No. NPF-8 for Joseph M. Farley Nuclear Plant, Units 1 and 2 (Accession No. ML013130226) is unchanged. The updated criticality safety analysis confirms that operation in accordance with the proposed amendment continues to meet the requ/red subcriticality margin.

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth In 1 O CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

E-12

)

SNC to NRC LAR Enclosure NL-19-0796

'4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the pr~posed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5. ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment

6. REFERENCES 6.1. NEI 12-16, Rev. 3, gGuidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants", March 2018.

6.2. K.Wood, "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools," DSS-ISG-2010-001, Accession Number ML102220567, Nuclear Regulatory Commission, Rockville, MD, August 2010.

I E-13

SNC to NRG LAR Enclosure NL-19-0796 ENCLOSURE Existing Technical Specification Page Markups Pages:

3.7.15-1 3.7.15-2 4.0-2 4.0-5 4.0-6 4.0-7 4.0-8 4.0-9 4.0-10 Insert 1 for 4.0-5 (7 Pages)

E-14

SNC to NRC LAR Enclosure NL-19-0796 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Assembly Storage Spent Fuel Assembly Storage 3-7.15 LCO 3.7.15 The combination of initial enrichment and bumup of each spent fuel assembly stored in the spent fuel storage pool shall be within the Aeeeptable BHmtJp Domain of FigtJfe 3.7.1 S--1 or in accordance with Specification 4.3.1.1.

APPLICABILITY:

Whenever any fuel assembly is stored in the spent fuel storage pool.

ACTIONS CONDl'TlON A.

Requirements of the LCO not met.

A.1 REQUIREO ACTION

~~~-~

~"' OTE~~~

l CO 3.0.3 is not applicable.

COMPLETION TIME Initiate action to move the Immediately noncomplying fuel assembly to an acceptable storage location.

SURVEILLANCE REQUIREMENTS SR 3.7.15.1 SURVEILLANCE Verify by administrative means the initial enrichment and bumup of the fuel assembly is in accordance with Figure 3.7.1 5 1 er Specification 4.3.1.1.

FREQUENCY Within 7 days following the relocation or addition of fuel assemblies to the spent fuel storage pool.

Farley Units 1 and 2 3.7.15-1 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

SNC to NRC LAR Enclosure NL-19-0796 45000 400* -,

'\\

35000

' '\\

30000 i=

~

I 25000

a.

=*

E ci

~ -

.0 20000 J '

15000 10000 5000 0

1.0

!Delete FigureJ.7.15-1. I

~

CEPTABLE

'\\.

\\

I\\.

\\

'\\

' \\.

\\

\\.

'\\

'\\.

I,.

I I

I,.

~

r

..i I

2.0 3.0 I

I r

'\\

ti..

Spent Fuel Assembly Storage 3 7 15 II I

.j 7T I

'\\

UNACCEfTABLE -....._

f\\

' "' ""\\.

"' '\\.

4.0

.0 Initial U-235 Enrichment {nominal w/o)

Figure 3. 7.15-1 Fuel Assembly Bumup limit Requirements For All Cell Storage Farley Units 1 and 2 3.7.15-2 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

SNC to NRC LAR Enclosure NL-19-0796 Design f eatures 4.0 4.0 DESIGN FEATURES 4.3.1.1 (continued)

b.

C.

ken < 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 4.3.2.7-2 of the FSAR; k~ s 0.95 if fully flooded with water borated to 400 ppm, which indudes an allowance for uncertainties and biases as described in Section 4.3.2. 7.2 of the FSAR; New or partially spent fuel assemblies that must be stored acoording to their com

• ation of discharge bulTl.lp and nominal emichmen decay time since operation, required Integral Fuel Bumab e Absorber (IFBA) Of applicable), d, and must comply with Figure 4.3-1, A nominal 10.75 inch center to center distance between fuel assemblies placed in the fuel storage racks; Table 4.3-1, and Tables 4.3-3 through 4.3-5 (as applicable). Each assembly should be stored in an appropriate storage configuration according to its fuel category as specifically described in Table 4.3-1 and geometry based on Figure 4.3-1; Fuel assemblies that are stored in accOfdance with every applicable storage array as shown in Figure 4.3-1 of which they are a part (i.e.,

one fuel assembly can be part of up to four different storage arrays, each storage array shall be in accordance with Figure 4.3-1); and

g.

New er partial~* speAt fuel assemblies with a eeml3iRatieA of disdia~e b1:1rn1:1p aRd initial eRriehmeRt iA the *aeeeptal31e Faflge" ef fig1:1FC 3.7.15 1 may be allowee t:.1RreslAeteEl storage in the speAt fuel sterage raek(s) (alse sho>++JR as the All Cell Storage coofig1:1ratioA iR Figure 4.3 2);

New or partially speAt fuel assemblies with a eombiRatioA of diseharge b1:1mup ooa iRitial eRrichment iA the *uAacceptable raRge" ef Figure 3.7.15 1 will be stereo in compliance With the NRG appl'O't'ed Figures 4.3-1 through 4.3-5. The high eAridlff'leRt fuel asseff'lblies showA in U\\e BttFAeEilFFC5A Sterage 00Rfig1:1ratfoA iA Figttre 4.3 2, with ff'I.Heiff'lt:.1ff'I AemiAal eAridlff'lents > 3.9 weight pereeAt U 235, shall eoAtain slifficieRl integral 131:1mable absoFbers sttch that a maxiff'l1:1m refereRee fuel assembly K.. ~ 1.455 at 68°F is ff'laiRtaiRml; aREl Unit 1 only -

Damaged fuel assemblies F02, F05, F06, F15, FH, F18, f 1,9, F20, F30, F31, and F32 shall be stored in accordance with Fig1:1re 43 6.~

4_3_ 1 _2 The new fuel pit storage racks are designed and shall be maintained with:

Far1ey Units 1 and 2

a.

Fuel assemblies with Standard Fuel Assembly fuel rod diameters having a maximum nominal U-235 enrichment of 4.25 weight percent; 4.0-2 (continued)

Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

SNC to NRC LAR Enclosure NL-19-0796 5

t--

~

0

~ -

a.

E

(])

2:-

.0 E

~

ti) qi u..

I 45000 400

'\\

35000 30000 25000 20000 15000 10000 5000 0

1.0 Delete Figure4.3-1 and insert revised Figure4.3-1 and Tables 4 3-1 throogh 4 3-5 (Insert 1)

\\.

/,

\\.

ACCEPTABLE J

I/

"\\.

/

\\.

/

\\.

~

\\.

/

J

\\.

I

' \\.

I b{

\\.

I

\\.

I

/

'\\

r,..

I

/,

Design Features 4.0 f,

f I/

J f

/,

~ NACCEPTABLE = -

~

I

"\\.

If

' "\\.

I I

'\\.

I

' \\.

I

\\.

I

' "\\.

2.0 3.0 4.0 0

tnruaJ U-235 Enrichment (nominal wfo)

Figure 4.3-1 Fuel Assembly Bumup Limit Requrrements for Low Ennchment (L)

Assembly of the Burned/Fresh Checkerboard Storage (see Figure 4.3-2)

Farley Units 1 and 2 4.0-5 Amendment No. 169 (Unit 1) I Amendment No. 161 (Unit 2)

SNC to NRC LAR Enclosure NL-19-0796

!Delete Figure 4.3-2 A

A All Cell Storage C

Empty Empty C

2-out-of-4 Storage Note:

A = All Cell Enrichmen (Figure 3.7. 1 ~ 1)

A A

H C = 2-out-of-4 Enrichment (No restriction on enrichment or bumup)

L = Low Enrichment of Burned/Fresh (Figure 4.3-1 )

Design Features 4.0 L

L H = High Enrichment of Burned/Fresh (See section 4.3.1.1 J for IFBA requirement)

Empty = Empty Cell Far1ey Units 1 and 2 Figure 4.3-2 Spent Fuel Storage Configurations 4.0-6 Amendment No. 169 (Unit 1) I Amendment No. 161 (Unit 2)

SNC to NRC LAR Enclosure NL-19-0796 Interface Note:

A A

L L

A = All Cell Enrichment jDele e Figure 4.3-3. j A

A A

A A

A L

L L

L l = Low Enrichment of Burned/Fresh H = High Enrichment of Burned/Fresh A

A A

A A

A A

A A

A A

A Boundary Between All Cell Storage and Burned/Fresh Storage Note:

Design Features 4.0 A

A A

A A

A

1. A row of empty cells can be used at the interface to separate the confi

2. It is acceptable to replace an assembly with an empty cell.

Farley Units 1 and 2 Figure 4.3-3 Interface Requirements 4.0-7 Amendment No. 169 (Unit 1) I Amendment No. 161 (Unit 2)

SNC to NRG LAR Enclosure NL-19-0796 Interface

~

Note:

A = All Cell Enrichment C = 2-out-Of-4 Enrichment Empty = Empty Cell

!Delete Figure 4.3-4_ I A

A A

A A

A Empty A Empty C

C Empty C A

A A

A A

A A

A Empty A A

A Design f eatures 4_0 A

A A

A A

A Boundary Between All Cell Storage and 2-out-of-4 Storage Note:

1 _ A row of empty cells can be used at the interface to separate the configuration 2_ It is acceptable to replace an assembly with an empty celL Farley Units 1 and 2 Figure 4_3-4 Interface Requirements 4_0-8 Amendment No. 169 (Unit 1) I Amendment No. 161, (Unit 2)

SNC to NRC LAR Enclosure NL-19-0796 jDelete Figure 4.3-5. j Empty C C

Empty Interface

~

Empty C Empty H

Note:

C = 2-out-of-4 Enrichment L = Low Enrichment of Burned/Fresh H = High Enrichment of Burned/Fresh Empty = Empty Cell Empty C

Empty H

Empty H

C Empty Empty C C

Empty Empty C

C Empty Empty C

Boundary Between 2-out-of-4 Storage and Burned/Fresh Storage Note:

Design Features 4.0 C

Empty C

Empty C

Empty

1. A row of empty cells can be used at the interface to separate the configur

2. It is acceptable to replace an assembly with an empty celL Farley Units 1 and 2 Figure 4.3-5 Interface Requirements 4.0-9 Amendment No. 169 (Unit 1) I Amendment No. 161 (Unit 2)

SNC to NRC LAR Enclosure NL-19-0796 G'k..-"N

~~'7:'r'r',

1-r,.

7 TT TT

!Renumber Figure 4 -~

as figure 4.3-2. j f31 Empty f 18 F17 f15 F20 F05 F32 Design features 4.0

-.;;;r

'T" 'T"...,;"'r T'

Note: All Assemblies are 3.0 w/o mu nominal enrichment Farley Units 1 and 2

~

Figure 4-:J..G Damaged Fuel Assembly Configuration (Unit 1 Only) 4.0-10 Amendment No. 169 (Unit 1) I Amendment No. 161 (Unit 2)

SNC to NRC LAR Enclosure NL-19-0796 Insert 1

• Any 2x2 array of storage cells containing fuel shall comply with the requirements of Array A, Array B, or Array C, as applicable.

A. Fuel is divided into two Groups, based on Fuel Type (Standard Fuel Assembly (STD)/Robust Fuel Assembly (RFA) or Optimized Fuel Assembly (OFA)).

B. Arrays A, B and C designate the pattern of fuel which may be stored in any 2x2 Array.

C. Fuel Categories 1-4 are defined in Table 4.3-1.

Array A 1

X Two Category 1 assemblies with two empty storage locations The Category 1 fuel assemblies must only be face adjacent to an empty storage location.

X 1

Array B 4

4 L

One Category 2 assembly with three Category 4 assemblies.

4 2

Array C 3

3 Four Category 3 assemblies.

3 3

Notes:

1. Any storage array location designated for a fuel assembly may be replaced with non-fissile material.

2. Empty locations designated with an X must remain completely empty.

Figure 4.3-1 Spent Fuel Pool Loading Restrictions Page 1 of 3

SNC to NRC LAR Enclosure NL-19-0796 Notes Continued:

Insert 1 (Continued)

3. Other Fuel Categories are determined as follows:

l

a. For STD/RFA assemblies, detem,ine the fitting coefficients A1 -A,. using Table 4.3-

3.

b. For OFA assemblies, determine the fitting coefficient~ A1 - ~ using Table 4.3-4.

c. For assemblies with Initial Enrichment (En) values greater than or equal to the values in Table 4.3-2, the required Minimum Bumup value (in MWd/MTU) for each Fuel Category is calculated based on initial enrichment, decay time, and the appropriate fitting coefficients. If the fuel assembly bumup is greater than the calculated Minimum Bumup value, then the fuel may be classified into this Fuel Category.

The equation for Minimum Bumup is:

Minimum Bumup (MWd/MTU) = 1,000 x [A1 x En3 + A2 x En2 + AJ x En + A,.]

./

Note: If the computed Minimum Bumup value is negative, zero shall be used.

The equation for Minimum IFBA required for Fuel Category 2 assemblies as a function of enrichment between 3.2 and 5.0 weight percent Uranium-235 is:

Minimum IFBA (rods) = A1 x En2 + Ai x En + AJ Note: The Minimum IFBA should be rounded up to the next whole number.

Note: Below 3'.2 weight percent U-235, IFBA is not required.

d. Decay time is measured in years. For decay times between the values in Tables, 4.3-3 and 4.3-4, linear interpolation or the lower decay time value may be used. If interpolation is used, linear interpolation based on actual decay time should be performed between calculated values of Minimum Bumup associated with tabulated Decay Times greater and less than the actual Decay Time. No extrapolation beyond 20 years is permitted.

e. Initial enrichment (En) is the nominal U-235 enrichment of the central zone region of fuel, excluding axial blankets. If the fuel assembly contains axial regions with different U-235 enrichment values, such as axial blankets,.the maximum enrichment value should be utilized. If the computed Minimum Burnup value is negative, zero shall be used.

Figure 4.3-1 Spent Fuel Pool Loading Restrictions Page 2 of 3

SNC to NRC LAR Enclosure NL-19-0796 Notes Continued:

Insert 1 (Continued)

4. An empty (water-filled) cell may be substituted for any fuel-containing cell in all storage arrays.

5. Fuel Category 2 fuel which has been operated must have at least 10,000 MWd/MTU of bumup.

Figure 4.3-1 Spent Fuel Pool Loading Restrictions Page 3 of 3

SNC to NRC LAR Enclosure NL-19-0796 Insert 1 (Continued)

Table 4.3-1 Fuel Categories Ranked by Reactivity Fuel Category 1 High ReactMty Fuel Category 2 Fuel Categq_ry 3 Fuel Category 4 Low Reactivity Notes:

1. Assembly storage is controlled through the storage arrays defined in Figure 4.3-1.

2. Fuel Categories are ranked in order of decreasing reactivity, e.g., Fuel Category 2 is less reactive than Fuel Category 1, etc.

3. Each storage cell in an array can only be populated with assemblies of the fuel category defined in the array definition or a lower reactMty fuel category.

4. Fuel Category 1 contains fuel with an initial maximum enrichment up to 5 weight percent U-235. Neither bumup nor IFBA is required.

5. Fuel Category 2 contains fuel with an initial maximum enrichment up to 5 weight percent U-235. Storage of fresh fuel is determined from the minimum IFBA equation and coefficients provided in Table 4.3-5. To be eligible for Fuel Category 2, fuel which has been operated in the reactor requires at least 10,000 MWd/MTU of bumup.

6. Fuel Categories 3 and 4 are determined from the minimum bumup equation and coefficients provided in Table 4.3-3 for STD/RFA fuel and in Table 4.3-4 for OFA fuel.

Table4.3-2 Maximum Enrichment allowed with 0.0 MWd/MTU Bumup Fuel Category RFA/STD OFA 1

5.0 5.0 2

5.01 5.01 3

2.15 2.15 4

1.70 1.75 Notes:

1. Requires I FSA credit for greater than 3.2 weight percent U-235.

2. For assemblies with an Initial Enrichment below the values listed above, no bumup is required

SNC to NRC LAR Enclosure NL-19-0796 Insert 1 (Continued)

Table 4.3-3 Coefficients to Calculate the Minimum Required Fuel Assembly Bumup (Bu) as a Function of Decay Time and Initial Enrichment (En) for STD/RFA Fuel Fuel Decay Time Coefficients Category (years)

A1 A2 AJ

~

0 0.2251

-2.5199 21.4065

-36.6115 5

0.3002

-3.4376 24.0978

-38.9002 3

10 0.1856

-2.3309 20.2704

-34 6503 15 0.0892

-1.3905 17.0683

-31.1550 20 0.0388

-0.9253 15.5082

-29.4500 0

-0.6112 4.6655 6.7127

-21.8911 5

--0.3326 2.0713 12.8468

-26.1880 4

10

--0.1305 0.0505 18.3242

-30.7080 15 0.1360

-2.6856 26.5239

-38.3300 20 0.2321

-3.7177 29.5977

-41.1200

SNC to NRC LAR Enclosure NL-19-0796 Insert 1 (Continued)

Table 4.3-4 Coeffi,cients to Calculate the Minimum Required Fuel Assembly Bumup (Bu) as a Function of Decay lime and Initial Enrichment (En) for OFA Fuel Fuel DecayTime Coefficients Category (years)

A1 A2 AJ

~

0 0.1692

-1.8852 18.5219

-32.7830 5

0.0191

-0.4154 13.4482

-27.1777 3

10

-0.0705 0.4300 10.5987

-24.0722 15

-0.1420 1.1146 8 2825

-21.5440 20

-0.1959 1.6375 6.5093

-19.6130 0

0.4957

-6.0715 37.2851

-49.1282 5

0.7476

-8.7581 45.3241

-56.5172 4

10 0.9041

-10.4334 50.3246

-61.0800 15 1.0799

-12 2326 55.7508

-66.1820 20 1.2541

-13.9154 60.5977

-70.5720

SNC to NRC LAR Enclosure NL-19-0796 Insert 1 (Continued)

Table 4.3-5 Fuel Category 2 Coefficients to Calculate the Minimum IFBA Required as a_

Function of IFBA Thickness and Fuel Type Coefficients Fuel Type IFBA Thickness A1

~

AJ 1.00X 5.2750 8.3325

-79.9546 STD/RFA 1.25X 3.7476 10.8046

-72.0974 1.50X 1.8593 19.8050

-81.5075 1.00X 6.2658 0.8890

-65.4949 OFA 1.25X 3.9144 9.3963

-68.9414 1.50X 1.5898 21.8436

-84.9630

SNC to NRG LAR Enclosure NL-19-0796 ENCLOSURE Retyped Technical Specification Pages

3..7 PlANT SYSTEMS 3.7_ 15 Spent Fuel Assembly storage Spent Fuel Assembly storage 3_7_15 LCO 3.7.15 The combination of lnfflal enlichmant and bumup of each spent fuel assembly stored m the spent fuel storage pool shall be ill accordance with

• Specifica1Ior1 4.3.1.1.

~ABIUTY:

. Whenever any fuel assembly Is stored l.n tne spent fuel storage pool_.

ACTHJNS.

CONDITION A.

Requirements of the LCO not met A.1 REQUIRED ACTION

---NOTE:---

LCO 3.0~3 is not~.

• COMPLETION TIME Initiate aclion to move the

_lmmediate1y noncomplynig fuel assembly to an acceptable mrage location.

SURVEILLANCE REQUIREMENTS SR 3.7.15.1 SURVEILLANCE Vefify by adm1nistrative means the 1nfflal enrichment and bumup *of the fuel assembly ts in accord~

with Specification 4.3. U.

FREQUENCY Within 7 days foDowilg the relocation or

• addlllon of fuel assemblies to the spent fuel stora~e pool Farley. Unifs 1 and 2 3.7.15-1 Amendment No.

(Unit 1)

'Amendment No.

(l!nl 2)

Design Features 4.0 4.0 DESIGN FEAlURES J

4.3.1.1 (contmled)

b. * ~< 1.0 ifUyfloooad with I.DX>f3.ted water. vmlcil Includes an allowance for unceJ'tan!les as described In Section 4.32.7 2 of theFsAR; c..

ke::r ~ 0..95 lf fully floorler1 wiH'I water ll orated to 400 p~ whlch itlcludes an ab.trance for uncertainties and biases as described in section 4.3.2.7.2 of the FSAR:

cl A nominal 10.75 mt center to centerorstance between fuel assemblies placed in 1fle Nel storage racks;

e.

• New or parflalfy spent fuel assembies that must be stored accordng to1hetr rombmallon of dischcll!Je bumup and nominru emichmem,. detaY time S10C8 operation, required Integral Fuel Bmiabte Absorher (IFBA) (rt' applcab{e}, and must com.ply wlh

~

4.3-1, TabJe 4.3-1, and Tablas 4.3-3_through 4_3-5 (as appIJcab!e). Each assembly should be stored in an approprtate storage configuratloo accordiOY to its~ categay as specifically descrlletf In Table 4.3-1 and geometry ~ased on Rgta'e 4.3-1; l

Fuel assembles 1flat are stored In acrordance wilt! ave,y app~e storage array as shown m FJgUm 4-_3-1 ofwhk:h they

. are a part [Le, one 1uel assembly can be part of up to four

• dilferent storage arrays, each storage array shall be In accordance with Fcgure 4.3-1 }; and

!J-Unit 1 on[y-Oamagecl fuel assembnes F02, FD5, F06, F15, F17, F18, F19, F20, F30, F31, and F32 shaU be stored in acconica1ce with Figura 4.3-2 4.3.1.2 The new fliel pll storage racks am designed and shat be maintained wtth:

a.

Fuel assembffles With standard Fua! Assembly fuel rod diameters havmg a maxtmum nominal U-235 emfchment of 425 weJght Percent.

(conlmued)

Faney Units 1 and 2 4.0-2 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

~Features 4.0 Arr,J 2x2 array of storage eels conmimng bllil shal comply wih"the reQUiremems of Array A, Array B, or Array c. as appilcah3e.

A. Fuel is divided lnlo two GrotJpS. based on Fuel 1}pe (Slandard Fuel Assembfy (SID)IRobust Fuel Assernh!y {RFA) or OJl(lnjzed Fueft Assembly (QFA))_ *

• R _ Arrays A. B and c* desJ@i:ate the pattern of l.Jel :whkh may be stored in ariy 2x2 Allay.

C. Fuel categ~ 1-4 are defined fn Table 4.3-1.

Amrv A I

1 X

~

Two Calegofy 1 assembles \\vial tw empt,; storage locaAJns.--The category 1

• fuel assembles must only be fare adJaced to an empty.storage location.

X 1

4 4

ArrayB One categmy 2 assembi'J with three ca&egory 4 asseimlies.

4 2

3, 3

An:avC Four Category 3 assemblfes.

3 3*

Notes:

1. Any storage array locaoon ~

for a fi.lel assemb]y may be replaced "'1th non:.tissrre material.

I

2. Empty locations designated wilh an X must rema11 complelely empty_

Farley: Unils 1 and 2..

Agura4~1 Spent Fuel Pool Loading Restiicfions Page 1 ofJ 4.D-5.

Amendment No_*

(Unit 1)

Amendment No.

(Unit 2)

3.. other Fuel Categories are detennined as~

. Design Featuras 4.0 a For SIDmFA assemblles, detem.*1e the tiffing coefficients A1 -A4 !,Ising Table 4..3-3.

b. For OFA assemllles, detemme the fflling coe1ficients At -A<< l.lSlng Table 4..3--4.

c.. For assemblies wiHl nlial Enrichment (En} values greater than or equal to the vaues in T~ 4..3--2. the mQUlred Mnimum Bumup V3Je (al M,WdlMTU) for each Fuel Category is caJcutded based on ~

emtchment, decay time, and the

. appropriate tiffing-ooeffldents_ H 1he rueJ assembly J)umup is greater than 1he cakuiated Minimum Bmlup:vaJue. fllen the fuel.lTlc)Y be classffied into. this Fuel categoly.

The equalion for Mmnmn Burnup is:

• 'Minlrrn.an Bumup {MWdiMTU) = 1,000 X [A1 X En3 + At. X En2 + A3 X En + Ail Note: If the computed M~ Bumup value m negaffVe, zero-shall be used The equalon forMirmnum I~ reQUI8d for Fuel ca~ 2 assemblies as a fundion of emk:hment balv.reen 32 and 5.0 ~:percent Uranium-235 is:

Mmlllll.lll lFBA (rods) = A1 X En2 + }Q X En + AJ Note: The fblirrun IFBA should be rounded up kl. the next whole number.

Note: Below 32 ~

percent U-235, IFBA Is not required.

d; Decay time Is measured In years_ Ford~ tsnes between th~ values In Tab1es 4..3-3 and 4_J-.4. ~lntefpolaflon ortha lower decay time ~ue ~ay be used. tt

• -intefpolation is* used, linear trrt8fJ)olatlon based on actµaJ decay time should be perfOJTJ"3d between calculated values of !Wnimum Bumup associated with tabulated Decay* Times greater and less than the aciUal Decay Tone. No extrapolation beyond 20 years is pennWed.

e. lntt:lal -enrfctunent (En} Is tfle :nomna1 U-235 enrkhment of the central zo~ region of fuel. exclrnmg axial blankels. lftll8 fuel assembly contains~ regions wtth -

differant U:~ ~ent ~~as axial blankets, the~ enrichment value shoukl be utllfzed. ~~

computed MinaTIU;ffi !3umup value is negative. zero shall be used.

Farley Units 1 and 2 Figure 4..3-1 Spent Fuel Pool Loading Restrictions Page2,of3 4.D-6 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

Nqtes Conmued:

Desjgn Fea1ures 4.0

4. An empty (Water-filed) ce!I may be stnitltuted for any fuel-containing cell m al storage arrays.

5. Fuel category 2 fuel 1,vhidl has been operated must have at least 10,000 MWdlMTU of

~-

Farley Units 1 and 2 Figure 4.3-1 Spent Fuel Pool Loadim) Restrictions Page 3of3 4.0-7 AmendmentNo.

(Unit 1)

Amendment No.

(Urut 2)

(

. \\

Table 4..3--1

• Fuel Categones R3:f1l$I by Reaclfvtt)t.

Fllel C3t8QWY 1 Hl1'lRea~

Fuel C3mgo:ry ~

Fuel category 3 Fuel~OJY4*

Low Reac(Mly

. Desfgn Featurns 4_0 NoiBs:

1. ~

stcragi, is controUed 1hrou9h 1tle storage arrays delned kl Agure 4_3-1_

2. Fuel CatBgori88 are ra1*6d in order of deCfflalWlg reactMty, e_g_,: Fuel Cafegoiy 2 is *ress ram:five than IFueJ C3tegOJy 1, ek.

3_ Each storage ceD !n en army' can only be pOJX*ded wflh assemblJes of the fuel~ defined in the mray definlion or a rower reactivlly fuel cmegory_

4. Fuel Cmegocy 1 *~

b!l 'Willl en aiHal maximum enrichnien1 up 1o 5 ~

percent U-235..

Nei!her bumup nor~ is required.

5_ *Fuel Camgo;y_ 2 co1111!d11s fuel~ mt Il!fi'11 mlIXfmum ~em up to 5 weight percent U-235_

storage ol' 1-esh fuel Is detennmed.i1>m Ille ninlmtn IFBA equation end coefficiemis priMded kl Table 4..3-5.. To be eligible to. Fuel Caegmy 2. fuel which ~

been operaled In Ille reeder

~

Id least 1D,OOO M\\IV~ of~-

6~ Fuel CafBgories 3 and 4 111B detmmfned from thD nmimwn bm1Up eqtmtion ood coefliclents,

proYilfed in Tabla 4..3-3 for STDIRFA fuel and In Table 4.3-4 for OFA 1iJa!..

Table 4..3--2 Maximum Enl1cylment ~

wtlti o_o ~~

Bwtq>

Fuel category RFNSTD OFJ\\

1 5_0.

,5_0.

2 5.01 5.01 3*

215 2.15 4

1-70

. 1.75 No!Bs:

1_ ~

IFBA credit for greater l!an'3.2 weight peretllJI: lf23S..

2. For BSS0ll1bies wHh an lnfflal Emf~ below~ valuss fisted

. ~. no bwnup Is requred Farfey Unils 1 arid ~

4.D-8 Am8Qdmant No.

(Unit 1)

Amendment No_

(Unit 2)

Table4.3-3 Design Features 4.0 Coefficferrts to C31cula'la the Mmimum. RequI8d Fuet Assembly ~

(Bu} as a Furaction of Decay Tune and lnfiial Enndlment {En) for STDIRFA Foo]

Fool DecayT'nne Coeflidenfs Catagm-y (years}

A1_*

A-i

~

A.

0 0.2251

-25199 21.4065

-36.6115 5

0.3002

-3.4376 241)978

-38..0002 3*

10 0.1856

-2.3309 202704

-34.6503 15 0.0892

-1.3905 17.1)683

-31.1550 20 0.0388

-0.9253

.. 15.5082

-29'.4500 0

-0.6112 4.6655 6.7127

-21.8911 5

-0.3326 2(f13*

128468

-26_1880 4

10

-0.1305

- 0.0505 18.3242

-30.7080 15

.0.1360

-26856 26.5239

-38.3300 20 02321

-3.7177 1 29.5977

-41.1200 Farley Units 1 and 2 4.0-9 Amendment No.

(Uni 1)

• AmendmentNo.

(Unit2)

• Table 4..3-4 Design Features 4.0 Coeticien1s to catcmde the Minimum Required Fuel Assembly Bumtip (Bu) as a Fmction of Decay Tine Md lnitia1 EIVidwnent (En} for OFA Fusi Fuel Decayl1me COefldems Calagoiy (yeaJs)

At A2.

A3 A4 0

0.1692

-1.8852 18.5219*

-32.7830 5

0.0191

-0.4154 13.4482

-T/.1777 3

10

-0.0705 0.4300 10.5987

-24.0722 15,

--0.1420 1.1146 8.2825

-21.5440 20

..0.1959 1.6375 6..5093

-19.6130 0

0.4957

-6.0715 37..2851

--49.1282 5

0.7476

.-8.7581 45.3241

-56.5172 4

10 0.9041

-10.4334 50.3246

-61.0800 15

'1.0799

-12.2326 55.7508

--66.1820 20 12541

-13.9154 60.5977

-70.5720 Farley Unis 1 and 2 4.0-10 Amendment No.

(lJJJit 1}

Amendment No.

(Unit 2)

' /

Table4.J.o Desl{Jn Features 4.0 Fuel categrny 2 Coefficienls to caJculate the Mnmun n=BA Reqfiied as a Ftmdlon of SFBA Thickness and Fuel lype Coel1cien;s Fuel Type.

IFBA Thickness A1

k.

A3 1.00X 52750 8..3325

-79.9546 STDIRFA 1..25X 3.7476 10.8046

-72...0974 1.50X 1.8593 19.8050

-81.5075 1.00X 62658 Q.8890

-65.4949 OFA 1.25X

.3.9144 9..3963

.:00.9414 1.50X 1..5898

• 71.8436

-84.~

Fartey Units 1 and 2 4.D-11 Amendment No.

(Unit 1)

Amendmant No.

(Unit 2)

._ TT"T T~

T TT T"7 F31 F18 F15 r

c

'IT Empty F17 F20

-_._ TT'TT"

,- "'.'l'"'I "'.IT F30 F06

)

F19 F32

'IT T

'"I.,..,,

Design Features 4.0

'I

"'T rx T

T Note: AJI Assemblies are 3:0 wlo 235lJ IU>flWlal enrichment Farfey Units 1 and 2 8gure4.3-2 Damaged Fuel Assembly Conl1guration

{Uni 1 Only) 4.0-12 Amendment No.

Amendment No.

(Uni 1)

(Unit 2)

SNC to NRC LAR Enclosure NL-19-0796 ENCLOSURE Existing Technical Specification Bases Page Markups Pages:

B3.7.14-1 B 3.7.14-2 B 3.7.14-3 B 3.7.14-4 B 3.7.15-1 B 3.7.15-2 B 3.7.15-3 B 3.7.15-4 I

SNC to NRC LAR Enclosure NL-19-0796 B 3-7 PLANT SYSTEMS Fuel Storage Pool Boron Concentra ion B 3I 14 B 3-7. 14 Fuel Storage 'Pool Boron Concentra -on BASES BACKGROUND

, decay time ssice operation, and req -Ted Integral Fuel Burnable Absorber (IFBA) [If applicable) comply with Figure 4.3-1, Ta le 4.3-1, and Tables 4_3-3 lllrough 4_3-5 (as applicable) of e Technical Speeifications.

• J_ M. Farley U its 1 and 2 Spent Fuel Pool Criticality Safety Analysis; WCAP-1841-hNP, Rev. 0 (Ret 4).

Figure 4_3-1, Table 4_3-1, and Tables 4_3-3 through 4-3-5.

Fuel assembr s.are stored in high density racks. The spent fuel s or.age racks contain storage locations for 1407 fuel assemblies.

Wes

• ghouse 17X17 fuel assemblies with initial enrichments less than or equal to 5.0 weight percent U-235 can be stored in any loca *on -n the spent fuel storage pool provided the fuel bumup-enrichment combinations ar:9 withiA tf:le limits speGified iR fi9~ira 3I1 5 1 af the Teetmieal Speeifieatians. Fuel assemblies that de Rat meet tf:le bumYp-GRriGhmeRt combiRatioR of Figure J.7.1a 1 may be stored irl tf:la speRt lYel stof:dge pool iR actordaAGe witf:I the patterns desclibed in Fi!}ures 4_3_1 1 threugh 4-3_1 5. The acceptable storage configurations are based on the "Westinghouse Spent Fuel Rack C micaliiy AAa!ysis M elhodology*, 'A'C.A,P 14 4 Hi NP A, Rel/_ 1, (Ref.

4) as iflplemeAted in the "Far1ey Units 1 and 2 Spent Fuel Rack Cffiicality P.flalysis Usiflg Saluele BaFeA Credit; CAA 97 138, Rev_ 1 (Ref. 7).

This n ethodology ensures that the spent fuel pool storage rack multiplication factor, Ke, is less than or equal to 0.95, as recommended by ANSI 57_2-1983 (Ref. 3) and NRC Guidance (Refs_

1, 2, and 6). A storage configuration is defined using l<eff calculations to ensure that Ke will be less than 1.0 with no soluble boron under normal storage conditions including tolerances and uncertainties_

Soluble boron credit is then used to maintain Ke less than or equal to 0.95. A spent fuel pool boron concentration of 400 ppm will ensure that l<eff will be less than or equal to 0_95 for all analyzed combina *ons of s orage patterns, enrichments, and bum ups. Th.e treatment of reactivity equivalencing uncertainties, as well as the calculation of postulated accidents crediting soluble boron is described in Ret4.

FicJure 4-3-1 and Table 4.3-1 _

The abo e methodology was used to evaluate storage f Westinghouse 17X17 fuel assemblies with initial enri ments less than or equal to 5.0 weight percent U-235 in the F spent fuel orage pooL The resul -ng enrichment and bum limlts are shown m Figure J.7.1 5 1 _ Checkerboard loading patte s are defined to allow storage of fuel assemblies that are not thin the acceptable Figure 4.3-1, Table burn up doma-n Figt.1Fe 3. 7.15 1. These s rage requirements are 4_3-1, and Tables4_3-3 L --~Mim*.necfi~aL Specification Fi91nes 4 _J _1 1 lf:lmugh 4.J.1 a. A through 4 _3-5_

(continued)

Farley Units 1 and 2 B 3-7.14-1 Revision O

SNC to NRC LAR Enclosure NL-19-0796 BASES BACKGROU 0 (con

• ued)

APPLICABLE SAFETY ANALYSES Farley Units 1 and 2 Fuel Storage Pool Boron Concentra ion B 3_7_14 spen fuel pool boron concentration of 2000 ppm ensures a no credible boron dilution event will result in a Ke greater an~o_g5_

E even damaged Westinghouse 17X17 fuel assemblies can in the Unit 1 spent fuel storage pool in the 12 storage cell configuration shown *n Technical Specifica *on Figure 4.3.1 6. The 11 fuel assemblies contain a nominal enrichment of 3.0 weight,percent U-235.

or cells Three accidents can be postulated for each storage co guration which could increase reactivity beyond the analyzed ndition. The three postulated accidents include a loss of the sp system, dropping a fuel assembly into an alread aded storage cell, aoo the misloading ot fuel assembly into a cell or which he restri

• cation, enrichment, or bumup are not sa *stied.

An increase in the temperature of the water passing through the s ored fuel assemblies causes a decrease in water density which would normally result in an addition of negative reacti ity. However, since Bora ex is not considered to be present in the criticality analysis, and the spent fuel pool water contains a high concentration of boron, a density decrease results in a positive reactivity addition.

The effect of an increase in reactivity due to an increase in emperature is bou~ th31rmisload accident.

~*

In the case of a fuel assembly dropped into an already loaded storage cell, the upward axial leakage of that cell will be reduced. However, the overall effect on the storage rack activity would be insignificant, since only the upward axial leakage of a single cell is minimized. ln addition, the neutronic coupling between the dropped fuel assembly and the already loaded assembly will be low due to a several inch separa *on of the active fuel regions due to the fuel assembly bottom no~le. The effects of this accident are also boun~

h~ misloa accident tora lo tio

~

~

s gecans The fuel assembly~i ding accident invo es the placement of assembly into for which the requirements on toca.tion, enrichment, or burnup are not met. This misload would result in a positive reactivity addition increasing *Ke toward 0.95. The amount of soluble boron required to compensate for the positive reactivity added is

, which is well below the LCO limit of 2000 ppm.

more U1an one fresh Sw/o U-235, unpoisoned 1710 ppm (continued)

B 3.7.14-2 RevisionO

SNC to NRC LAR Enclosure NL-19-0796 BASES APPLICABLE SAFETY ANALYSES (continued)

LCO APPLICABILITY ACTIONS Fariey Units 1 and 2 Fuel Storage Pool Boron Concentration B 3.7_ 14 A spent fuel pool boron dilution evaluation determined that the volume of water necessary to dilute the spent fuel pool from the l CO limit of 2000 ppm to 400 ppm (the boron concentration required to maintain l<etfless than or equal to 0_95) is approximately 480,000 gallons_ A spent fuel pool dilution of this volume is not a credible event, since it would require this large volume of water to be transferred from a source to the spent fuel pool, ultimately overflowing the pool. This event would be detected an~ terminated by plant personnel prior to exceeding a l<eff of 0_95_

The concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of 10 CFR 50_36(c)(2)(ii).

The fuel storage pool boron concentration is required to be

~ 2000 ppm. The specified concentration of dissolved boron in the fuel storage pool preserves the assumptions used in the analyses of the potential criticality accident scenarios as described in Reference 5. The specified boron concentration of 2000 ppm ensures that the spent fuel pool Kett will remain less than or equal to 0_95 due to a postulated fuel assembl isload accident (85Q ppm) or boron dilution event (400 ppm)_

mulliple

~

This LCO applies whenever fuel assemblies are stored in the spent fuel storage pool.

A.1 and A-2 The Required Actions are modified by a Note indicating that l CO 3_0_3 does not apply.

When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assembl"es. Action is also initiated to restore the concentration of boron simultaneously with suspending movement of fuel assemblies.

(continued) 8 3.7.14-3 Revision 0

SNC to NRC LAR Enclosure NL-19-0796 BASES ACTIO S SURVEILLANCE REQUIREMENTS REFERENCES WCAP-18414-NP, Rev. 0, *J.M.

Farley Units 1 and 2 Spent Fuel Pool Critica iy Safety Anatysis,"

September, 2019.

Fartey Units 1 and 2 A.1 and A.2 (continued)

Fuel Storage Pool Boron Concentration B 3.7.14 If the LCO is not met while moving irradiated fuel assemb~es in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movemenl is independent of reactor opera *on. Therefore, inability to suspend movement of fuel assemblies is no s cient reason to require a reactor shutdown.

SR 3.7.14.1 This SR verifies that the concentration of boron in the fuel storage pool is within the required limit. As long as this SR is met. the analyzed accidents are fully addressed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

1. USNRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, June, 1987.

2. USNRC Spent Fuel Storage Facility Design Bases (for Comment)

Proposed Revision 2, 1981.

3. ANS, "Design Requirements for *gh Water Reactor Spent Fuel Storage Facirities at Nuclear Power Sta ions," ANSUANS-57.2-1983.

4. WCAf? 14416 NP A, Rev. 1, "We8tin9hQYS8 Spent FYal RaGk Gntiealily Analysis Mcthee!olo!JY. NO\\lcml:lcr, 1996.

5. FSAR, Section 4.3.2.7.2.

6. NRC, Letter to all Power Reactor Licensees from BK Grimes,

" OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14, 1978.

1--: " Failey Units 1 aRd 2 Sp&Rt FY&[ ~ aGk ClitlGality Anal~ i8 Using Seluble Baron Creel it," GM 97 138, Re*.*- 1.

B 3.7.14-4 Revision 52

SNC to NRC LAR Enclosure NL-19-0796 B 3-7 PLANT SYSTEMS Spent Fuel Assembly Storage B 3.7.15 B 3.7.1 5 Spent Fuel Assembly Storage BASES BACKGROUND

, decay time since operation, and required Integral Fuel BIDlBble Absorber (IFBA} (if applicable) comply wrth Figure 4.3--1, Table 4.3-1, and Tables 4_3-3 through 4.3--5 (as apprcable) of tile Te.chnical Specifications.

• J_ M_ Farley Units, and 2 Spent Fuel Pool Criticality Safety Analysis,*

WCAP-18414-NP, Rev. 0 (Ret 1).

Westinghouse 17X17 fuel assemblies are stored acrording to their fuel type (Standartl 1Fuel Assembly (SID)/

Robust Fuel Assembly (RFA) or Optimized Fuel Assembly (OFA)),

decay time since last operation, nominal enrichment and bumup.

Westinghouse 17X17 fuel assemblies with nominal enrichments less than or equal to 5.0 weight percent U-235 can be stored in Array A as Fuel Category 1 as shown in Figure 4.3-1. In the Array A checkerboard storage arrangement, 2 fuel assemblies can be stored comer adjacent wrth empty storage cells.

Fuel assemblies are stored in high density racks_ The spent fuel storage racks contain storage tocalions for 1407 fuel assemblies.

Westinghouse 17X17 fuel assemblies with initial enrichments less than or equal to 5.0 weight percent U-235 can be stored in any location in the spent fuel storage pool provided the fuel bumup-enrichment combinations are wittiin U:le limits speGifieEI iR FigYre

3. 7.15 1 of !he Tee-hAieal S13ecifieations_ Fuel assemblies that do Flot meet the burnYp-enriGhment combiRalion of FigYre J.7.1 a 1 may be storeEI in ttie spent fuel swrage pool in aGGordance with the pattems destnbed ifl Fig1:JFes 4.3-1 1 lhre1:1gh 4.3-1 5. The acceptable storage configurations are based on the " Westiflgho1:1se Spcflt Fuel Ratk Ci:itiGality Analysis Methodology," WCAP 144 Hi NP A, Re*1. 1, (Ref.

1) as im13lemeRted in "Fafley LIA its 1 aAEl 2 SpeRt F1:1el Rack GRtitalityAAalysis Usiflg Sol1:1Blc 80R>A Credit," CM 97 138, Rev_ 1 (Ref. 2}.

The following storage configurations and enrichment limits were evaluated in the spent fuel rack criticality analysis:

WestiRgl=leuse 17X17 fuel assemo[ies \\Vilh nomiAal eArithmeRts less than or eq1ial lo 2_15 weight percent U 235 Gan be sloreEI in any cell locatieA as sho>ltR if Figure 4.3.1 2. Fl:fc I asscmtllies with iRitial AomiRal eRRcl=lments greater tl=laR these ~mits ml-1st satisfy a miAiml-lm bmnYp requirement as shown in Fi9Yre JJ.15 1.

lA'estiAghouse 17X17 fuel assemblies with nemiRal eArichments less than or 0q11al lo 5Jl weight percent *J 2Ja GaA be stored iR a 2 oYt of 4 checkerooard aR"angemeRt as shewA in Figure 4.3.1 2. IA the 2 out of 4 ehetkertrnar=e storage aFFangement, 2 fuel assemblies CilA be

---

,,- sloi:eEI Gomer aEljacent with empty stor:age Galls.

Westinghouse 17X17 fuel assemblies can be stored *n a burned/fresh arrangement (Army 8, Fuel Categories 4 and 2) of a 2X2 matrix of storage cells as shown in Figure4.3--1. In the ArrayB arrangement, assemblies must satisfy the minimum bumup and enrichment requirements fer fuel Category 4 assemblies as shown in figure 4.3-1, and Tables 4.3-1, 4.3-3, and 4.3-4, or Fuel Category 2 assemblies must meet the enrichment and If BA requirements of Figure 4_3-1 and Tables4.3-1 and 4.3--S.

WestiRghouse HX1'7 fuel assemblies ean tle sterce iR a tlulfleEl/frestt GAec:kerboard arr;ai:i;gemem: of a 2X2 matrix of storage Galls as shown in figure 4.J.1 ;i_ IR the tn.1meEl/t~eshi 2X2 Gilec:kemoard aFTa.RgemeRt, ttlree of the fuel assemblies must l:ia\\'e an iAitial Aominal enriGnmeRt less !Ran gr e~:Yal to 1.6 weighit per;cent U 2J5, or satisfy a minimYm b1,1m1ip reqYiremeAl for Ai9her ir:iitial er:iriGhmeRt8 as snown iR Fi!JUFe 4.3.1 1.

Westinghouse 17X17 fuel assemblies can be stored in a uniform "al-ceJr arrangement (Array C) of a 2X2 matrix of storage eels as shown in f igure 4.3-1. In the 1----------~

~ --iArray C all-cell arrangement, assemblies must satisfy B 3-7.15-1 the minimum bumup and enrichment requirements of Fuel Category J assemblies as shown in Figure 4.3-1 and Tables 4.3--1, 4.3-3, and 4.3-4.

SNC to NRC LAR Enclosure NL-19-0796 BASES BACKGROUND (continued)

\\

APPLICABLE SAFETY ANALYSES single fuel assembly oc mul pie fuel assemb es Farley Units 1 and 2 Spent Fuel Assembly Storage B 3.7.15 The fouit h fuel assembly must have an initial RomiRal eRr:iGhment less than or equal to J_g weight per:c~int U 2J5, or :&atisfy a miRimum lntegr:al Fuel Qumable A.bsomer rgquiremeRt for t:liQher initial eAFiehmoots te maintain the refcronee ruel assembly K-less than er e~t:Jal to 1.45§ at 68°F.

Eleven damaged Westinghouse 17X17 fuel. assemblies can be stored in the Unit 1 spent fuel storage pool in a 12 storage cell con guration surrounded by empty cells as shown in Technical Specification Figure 4:3:4--e-. The 11 fuel assemblies contain a nominal enrichment of 3.0 weight percent U-235.

Three accidents can be postulated for each storage c which could increase reactivity beyond the analyzed or-ce Is three postulated accidents include a loss of the sp t fuel pool cooling system, dropping a fuel assembty into an al read

• oaded storage cell, and the misloading of into a cell or which the restric

• a n, enrichment, or bumup are not satisfied.

An increase in the temperature of the water passing through the stored fuel assemblies causes a decrease in water density which would normally result *nan addition of negati e reactivity. However, since Bora ex is not considered to be present in the criticality analysis, and the spent fuel pool water contains a high concentration of boron, a density decrease results in a positive reactivity addition.

The effect of an increase in reac *vrty due to an tncrease in temperature is boun~

th~ misload acc*dent.

~ -

In the case of a fuel assembly dropped into an already loaded storage cell, the upward axial leakage of tha cell wm be reduced. However, the overall effect on the storage rack activity would be insignificant, since only the upward axial lea age of a single cell is minimized. In addition, the neutronic coupling between the dropped fuel assembly and the already loaded assembly "Cl be low due to a several inch separation of the active fuel regions due to the fuel assembly bottom no2:21e. The effects of this accident are also bou~ th~ misload accident.

~

(continued)

B 3.7.15-2 Revision O

SNC to 'NRC LAR Enclosure NL-19-0796 APPLICABLE SAFETY ANALYSES (continued)

LCO

, decay time, IFBA requirements, and/or btmup of lhe fuel assembly are specified

• n Figure 4.3-1, Table 4.3-1, and Tables 4.3-3 lhrough 4.3-5 for all spent fuel pool storage configllalions.

APPLICABILITY ACTIONS Farley Units 1 and 2 The fuel assembly misloading ace*

t involves the placement of a fuel assembly into a starage leeatien for which the requirements on location, enrichment, or bumup are not met. This misload would result in a pos* *ve reactivity addition increasing l<eff toward 0.95. The amount of soluble boron required to compensate for the positive reactivity added,is

, which is well below the LCO limit of 2000 ppm.

1710 ppm The configuration of fuel assemblies in the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

The restrictions on the placement of fuel assemblies within the spent fuel pool ensure the Kerr of the spent fuel storage pool will always remain < 0.95, assuming the pool to be flooded with borated water_

The combination of initial enrichment and bumup are speeified in f igure J. 7.1 6 1 for the J\\ 11 Cell Ste rage C0Afi91;.11=atian. Other aeeept-able emiel:1ment. bumup, and el:leekerbaaFEl starage oonfiguratiaAS are speeified in f igures 4.J.1 1 through 4.J.1 6.

This LCO applies whenever any fuel assembly is stored in the spent fuel storage pool.

Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When the configuration of fuel assemblies stored in the spent fuel storage pool is not in accordance with the acceptable combination of initial enrichments, bumup, and storage configurations, the immediate action is to initiate action to make the necessary fuel assembly movement(s} to bring the configuration into compliance with f igureJ.7.15 1 ar Specification4.3.1.1.

If unable to move irradiated fuel assemblies wh~e in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

B3.7.1 5-3 Revision 0

SNC to NRG LAR Enclosure NL-19-0796 BASES SURVEILLANCE REQUIREMENTS

, decay -me, FBA req -rements, amtfor-b mup o the fuel assembty is

• in e ccep e bumup domail of Figw-e 4_3-1, Table 4.3-1, and Tables 4.3-3 through 4_3-5_

WCAP-18414-NP, Rev. 0, *J.M.

Farley Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis

• September, 2019.

REFERENCES Farley Units 1 and 2 SR 3-7.15.1 Spent Fuel Assembly Storage B 3.1_15 This SR e

• es by administra - e means (e.g., Core Loading Plan, ote computer code output or TrackWorks program) that the initial enrichmen and bumup gftt:le ruel,uosembfy is within tl=le aGGeptable bumuµ dor:llaiR gfFiQblr:9 J.7.15-1. Fgrruel assemblies in lt:le 1,maeee~lilble faAge af FiglHe 3.7.1§ 1, ~erfeffflaAee eftl=tis SR will also ensur:Q compliance witl=I Specification 4.J.1.1.

The frequency of within 7 days follow-ng the relocation or addi *on of fuel assemblies to the spent fuel storage pool ensures that fuel assembres are stored within the configuration analyzed in the spent fuel rack criticality analysis. Th.is surveillance would be performed after all of the fuel handling is completed during a refueling outage, or new ruel assemblies are placed into the spent fuel pool.

his SR does not have to be performed following interruptions in fuel handling: during defined fuel movements as described above (i.e., it is only required after all fuel movement associated with refueling operations is completed) or if only certain fuel assemblies are reloca ed to d-eren storage locations within the pool (onty the moved assembres ust be verified).

The 7 day al owance for completion of this Surveillance provides adequate time for completion of a spent fuel pool inventory verification while mtnlmizing the time that a fuel assembly could be misloaded during a refueting or the placement of new fuel assemblies into the spen fuel pool. The boron concentration required by Specification 3.7.14 ensures that the spent fuel rack Ke remains within limits until the spent fuel pool *nventory verification is performed.

1. \\*VG,"iP 14416 NP A, Re\\'. 1, " WestiA§R81:1Se SpeAl F1:1el Rack Ci:itic;:ality Anal~is Metf:lggglggy, NQYember, 1QQ6.

2. *Fooey UAits 1 aRd 2 S~eflt F1:1el Raek Gntiealily Aflalysis UsiAg SQIYble BQFQR Cr:edit; C.AA g7 13B, Re\\'_ 1.

B 3.7.15-4 Revision O

SNC to NRC LAR Enclosure NL-19-0796 ENCLOSURE Westinghouse Application for With holding Proprietary Information from Public Dlsclosure CAW-19-4943, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice

Westinghouse Non-Proprietary Class 3 AFFIDAVIT COMMONWEAL TH OF PENNSYLVANIA:

COUNTY OF BUTLER:

(1)

I, Camille T. Zozula, have been specifically delegated and authorized to apply for CAW-19-4943 Page 1 of 3 withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2)

I am requesting the proprietary portions ofWCAP-18414-P, "J.M. Farley Units 1 & 2 Spent Fuel Pool Criticality Safety Analysis" be withheld from public disclosure under 10 CFR 2.390.

(3)

I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4)

Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i)

The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii}

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitprs to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses.

Also, public disclosure of the information would enable others to use the information.

to meet NRC requirements for licensing docwnentation without purchasing the right to use the information.

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' T Westinghouse Non-Proprietary Class 3 CA W-19-4943 Page 2 of 3 AFFIDAVIT (5)

Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage,' as follows:

( a)

The information reveals the distinguishing aspects of a process ( or I

component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b)

• It consists of supporting data, including test data, relative to a process ( or component, structure, tool, method, etc.), the application of which data secures a competitive ecciriomk advantage (e.g., by optimiz.ation or improvecr- -

marketability). '

( c)

Its use by a competitor would reduce his expenditure of resources or improve (d)

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(e) his competitive position in the design, manufacture, shipment, ~on, assurance of quality, or licensing a similar product.f It reveals cost ot price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f)

It contains patentable ideas, for which patent protection may be desirable.

(6)

The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding 1s indicated in both versions by means oflower case letters (a) through (t) located as a superscript immediately f9llowing the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters

)

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Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW:...19-4943 Page 3 of3

refer to the types of.information Westinghouse customarily holds in confidence identified in Sections (5Xa) through (f) ofthis Affidavit I declare that tµe averments of fact set forth in this Affidavit are true and correct.to the best ofmy kno~ledge, information, and belief.

I declare~ penalty of perjury that the foregoing i~ true and correct 4-A'"l:f-lJl<L, Manager

• oactive Materials I

• 1 !

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• l

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and non-proprietary versions of a document, furnished to the NRC in connection with requests for plant-specific review and approval.

In order to conform to the requirements of IO CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted).

The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (5)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b )(I).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a.

license, permit, order, or regulation subject to the requirements of IO CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to th'e non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.