RS-23-093, License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel.

License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel.

ML23272A201
Person / Time
Site:	Byron, Braidwood
Issue date:	09/29/2023
From:	Lueshen K Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML23272A200	List: RS-23-093, License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel.
References
RS-23-093
Download: ML23272A201 (1)

Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office Proprietary Information - Withhold from Public Disclosure Under 10 CFR 2.390

RS-23-093 10 CFR 50.90 September 29, 2023

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457

Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

Subject:

License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, "Spent Fuel Pool Boron Concentration", 3.7.16, "Spent Fuel Assembly Storage, 4.3.1 "Fuel Storage, Criticality"

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit, " Constellation Energy Generation, LLC (CEG) requests amendments to the Technical Specifications (TS) for Braidwood Station, Units 1 and 2 (Braidwood) and Byron Station, Units 1 and 2 (Byron).

The proposed amendment is to change:

x Technical Specifications (TS) 3.7.15, "Spent Fuel Pool Boron Concentration" to increase the required spent fuel pool boron concentration to be > 2000 ppm.

x TS 3.7.16, "Spent Fuel Assembly Storage" to update Figure 3.7.16-1 to include fuel from Framatome and Westinghouse.

x TS 4.3.1.b, "Fuel Storage," "Criticality" replaced with keff < 1.00, at a 95% probability, 95% confidence level, if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1.2.3.1 of the Updated Final Safety Analysis Report (UFSAR) and keff < 0.95, at a 95% probability, 95% confidence level, if fully flooded with water borated to 550 ppm, which includes allowances for biases and uncertainties as described in Sections 9.1.2.3.5 and 9.1.3.1 of the UFSAR.

x TS 4.3.1.c and d, "Fuel Storage," "Criticality" replace "For Holtec spent fuel pool storage racks, a" with "A". (Braidwood only)

CEG performed a criticality safety evaluation for fuel assembly storage in the Braidwood and Byron Spent Fuel Pool (SFP) storage racks and New Fuel Storage Vaults to support the proposed TS change using a methodology described in Attachments 1 and 8.

Attachment 8 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachment 8 this document is decontrolled.

September 29,2023 U.S. Nuclear Regulatory Commission Page 2

The attached amendment request is subdivided as follows:

x Attachment 1 provides an evaluation of the proposed change. (Non-Proprietary) x Attachment 2 provides the current Braidwood TS pages with the proposed change indicated with mark-ups.

x Attachment 3 provides the current Byron TS pages with the proposed change indicated with mark-ups.

x Attachment 4 Affidavit CEG.

x Attachment 5 Affidavit Framatome.

x Attachment 6 provides NEI 12-16 checklists.

x Attachment 7 Summary of Regulatory Commitments x Attachment 8 provides an evaluation of the proposed change. (Proprietary) contains information proprietary to CEG and Framatome. Each affidavit, provided in Attachments 4 and 5, respectively, sets forth the basis on which the corporations information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, it is respectfully requested that the information that is proprietary to CEG and Framatome be withheld from public disclosure in accordance with 10 CFR 2.390. Attachment 1 provides the non-proprietary version of the document provided in.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), CEG is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

CEG requests approval of the proposed license amendment by September 30, 2024. Once approved, the amendment will be implemented within 90 days.

The proposed amendment has been reviewed and approved by the Braidwood and Byron Plant Operations Review Committee in accordance with the requirements of the CEG Quality Assurance Program.

There are regulatory commitments contained in this letter. They are described in Attachment 7.

Should you have any questions concerning this letter, please contact Ms. Lisa Zurawski at (779) 231-6796.

Attachment 8 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachment 8 this document is decontrolled.

September 29,2023 U.S. Nuclear Regulatory Commission Page 3

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 29th day of September 2023.

Respectfully, Lueshen, Digitally signed by Lueshen, Kevin Kevin Date: 2023.09.29 10:19:33 -05'00' Kevin Lueshen Sr. Manager - Licensing Constellation Energy Generation, LLC

Attachments:

1. Evaluation of Proposed Change (Non-Proprietary)

2. Braidwood Mark-up of Technical Specifications Page

3. Byron Mark-up of Technical Specifications Page

4. Affidavit CEG

5. Affidavit Framatome

6. NEI 12-16 Checklists

7. Evaluation of Proposed Change (Proprietary)

cc:

NRC Regional Administrator - Region III NRC Senior Resident Inspector - Braidwood Station NRC Senior Resident Inspector - Byron Station Illinois Emergency Management Agency - Division of Nuclear Safety

Attachment 8 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachment 8 this document is decontrolled.

ATTACHMENT 1

Evaluation of Proposed Change (Non-Proprietary)

In this attachment, Framatome information is indicated by the use of a single brackets (e.g., []) as described in Attachment 5 and Constellation Energy Generation, LLC information is indicated using a double brackets (e.g., (())) as described in Attachment 4.

ATTACHMENT 1 Evaluation of Proposed Change

Subject:

License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, "Spent Fuel Pool Boron Concentration", 3.7.16, "Spent Fuel Assembly Storage, 4.3.1 "Fuel Storage, Criticality"

1.0

SUMMARY

DESCRIPTION

2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

4.2 Precedents

4.3 No Significant Hazards Consideration Determination

4.4 Conclusions

6.0 REFERENCES

Page 1 of 54 ATTACHMENT 1 Evaluation of Proposed Change

1.0

SUMMARY

DESCRIPTION

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit, " Constellation Energy Generation, LLC (CEG) requests amendments to the Technical Specifications (TS) for Braidwood Station, Units 1 and 2 (Braidwood) and Byron Station, Units 1 and 2 (Byron).

The proposed amendment is to change:

x Technical Specifications (TS) 3.7.15, "Spent Fuel Pool Boron Concentration" to increase the required spent fuel pool boron concentration to be > 2000 ppm.

x TS 3.7.16, "Spent Fuel Assembly Storage" to update Figure 3.7.16-1 to include fuel from Framatome and Westinghouse.

x TS 4.3.1.b, "Fuel Storage," "Criticality" replaced with keff < 1.00, at a 95% probability, 95%

confidence level, if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1.2.3.1 of the Updated Final Safety Analysis Report (UFSAR) and keff < 0.95, at a 95% probability, 95% confidence level, if fully flooded with water borated to 550 ppm, which includes allowances for biases and uncertainties as described in Sections 9.1.2.3.5 and 9.1.3.1 of the UFSAR.

x TS 4.3.1.c and d, "Fuel Storage," "Criticality" replace "For Holtec spent fuel pool storage racks, a" with "A". (Braidwood only)

CEG performed a criticality safety evaluation for GAIA fuel assembly storage in the Braidwood and Byron Spent Fuel Pool (SFP) storage racks and New Fuel Storage Vaults (NFV) to support the proposed TS change using a methodology described in Attachments 1 and 8.

The criticality safety analysis of record (Reference 6.1.9) for the SFP will still cover Westinghouse supplied 17x17 fuel assemblies (OFA (Optimized Fuel Assembly), VANTAGE 5, VANTAGE+) and previously irradiated Westinghouse and Framatome Lead Use Assemblies (LUAs). The enrichment limit for Westinghouse supplied 17x17 fuel assemblies (OFA, VANTAGE 5, VANTAGE+) and Framatome LUAs is 5.0 weight percent (wt%) Uranium 235 (U-235).

Page 2 of 54 ATTACHMENT 1 Evaluation of Proposed Change

2.0 DETAILED DESCRIPTION

2.1 Proposed Changes

The proposed TS changes are provided in Attachments 2 and 3 and shown in the table below:

Current TS 3.7.15 Proposed TS 3.7.15 The spent fuel pool boron concentration shall The spent fuel pool boron concentration shall be > 300 ppm. be > 2000 ppm.

Current TS Figure 3.7.16-1 Proposed TS Figure 3.7.16-1 Figure shows Region 2 Fuel Assembly Figure changed to reflect Region 2 Fuel Burnup Requirements for Westinghouse fuel. Assembly Burnup Requirements for Westinghouse and Framatome fuel.

Current TS 4.3.1.b Proposed TS 4.3.1.b

b. For Holtec spent fuel pool storage racks, b. keff < 1.00, at a 95% probability, 95%

keff < 0.95 is fully flooded with unborated confidence level, if fully flooded with water, which includes an allowance for unborated water, which includes an uncertainties as described in Holtec allowance for biases and uncertainties as International Report HI 982094, "Criticality described in Section 9.1.2.3.1 of the Updated Analysis for Byron / Braidwood Rack Final Safety Analysis Report (UFSAR) and Installation Project," Project No. 80944, 1998. keff < 0.95, at a 95% probability, 95%

confidence level, if fully flooded with water borated to 550 ppm, which includes allowances for biases and uncertainties as described in Sections 9.1.2.3.5 and 9.1.3.1 of the UFSAR.

Current TS 4.3.1.c and TS 4.3.1.d Proposed TS TS 4.3.1.c and TS 4.3.1.d (Braidwood Only) (Braidwood Only)

For Holtec spent fuel pool storage racks, a A

The Byron/Braidwood Updated Final Safety Analysis Report (UFSAR) will be updated in accordance with 10 CFR 50.71(e) as part of implementation of the approved amendment. A summary of the proposed changes is provided below.

x Section 9.1.1, "New Fuel Storage", will be modified to reflect storage requirements of GAIA fuel in the NFV.

x Section 9.1.1.3, "Safety Evaluation", will be revised to reflect the NFV requirements of 10 CFR 50.68(b)(2) in this section.

x Section 9.1.2, "Spent Fuel Storage", and 9.1.2.3, "Safety Evaluation" will be updated to reflect the characteristics of the new SFP criticality safety analysis (CSA) covering GAIA fuel.

x Section 9.1.2.3.10, "Fuel Rack Design Features", and 9.1.6, "References", will be updated for consistency with other changes in Section 9.1.

x Section 9.1.3.1, "Spent Fuel Pool Cooling and Cleanup System", "Design Bases", will be updated for consistency with other changes in Section 9.1.

Page 3 of 54 ATTACHMENT 1 Evaluation of Proposed Change

2.2 Background

2.2.1 New Fuel Vault

The New Fuel Vault (NFV) provides dry storage for 132 fresh fuel assemblies, and it is made of three rows of 44 stainless steel fuel storage cells. The NFV exists in a reinforced concrete pit designed to prevent flooding, but it is nonetheless analyzed for criticality concerns in the event of water intrusion.

2.2.2 Spent Fuel Pool

The Spent Fuel Pool (SFP) consists of two distinct zones to safely store fuel assemblies: Region 1 and Region 2. These storage locations allow for effective cooling and decay time of assemblies until they are suitable for relocation to a Dry Cask Storage (DCS) system and/or an eventual permanent spent fuel repository. The Region 1 cells can accommodate both fresh and spent fuel, while the Region 2 cells can only accommodate fuel assemblies with minimum discharged burnups.

Region 1 is significantly smaller at 396 cells, compared to the Region 2 capacity of 2588/2568 cells at Byron and Braidwood respectively.

Region 1 racks are composed of SA240-304L stainless steel sheets and plates, and SA564-630 stainless steel bar material. On each wall of a fuel cell there is a neutron absorber composed of Boral to suppress criticality. Between inner boxes a gap element is welded in place to serve as a neutron flux trap in addition to structural benefits. This flux trap functions by allowing more opportunity for moderation and thermalization, as the Boral inserts will more readily absorb a thermal neutron.

Region 2 racks are composed of the same materials as Region 1 racks, however the geometry was altered. The overall area of each cell was reduced by 0.07 in2 and the flux trap has been eliminated. This lack of flux trap and closer geometry requires the use of a burnup-enrichment acceptance criteria curve in order to load assemblies; this curve is unique to Region 2.

The SFP criticality requirements were most recently altered via Amendment No. 112/105 (Ref.

6.1.1) to Byron and Braidwood stations, respectively. Several of the alterations proposed in this request will change Byron and Braidwoods Technical Specifications, similar to those approved in Amendment No. 94/86 (Ref. 6.1.8).

Currently, for normal conditions, the design basis for preventing criticality of Regions 1 and 2 is based on an effective neutron multiplication factor, keff, including uncertainties and biases to be less than or equal to 0.95 with a 95% probability and 95% confidence level with the racks fully loaded with the highest anticipated reactivity assemblies. This is in accordance with ANSI 57.2-1983 and NRC Guidelines (Refs. 6.1.2 and 6.1.3). Conservatively, the SFP is assumed to contain clean, un-borated water at the maximum water density (4 °C).

Page 4 of 54 ATTACHMENT 1 Evaluation of Proposed Change

2.2.3 Fuel

The existing analyzed fuel types at Byron and Braidwood are Westinghouse Vantage 5, Vantage 5+, and Optimized Fuel Assembly (OFA). ((

)), all future reload batches will utilize Framatome GAIA fuel assemblies. The initial reloads of the GAIA fuel type will be of the "baseline" 18-month cycle variety. CEG intends to eventually load fuel with a 24-month runtime capacity. To support that intent, the GAIA assembly parameters used in the SFP and NFV criticality safety analyses are designed to encompass the 24-month cycle GAIA bundle with higher enrichment and burnup capabilities.

The GAIA fuel is designed for Westinghouse-type plants with a 17x17 fuel rod array. The cladding material is M5, a Framatome advanced zirconium alloy. The design utilizes a 144-inch fuel stack length of UO2 or UO2-Gd2O3 (Gadolinia).

In this License Amendment Request (LAR), the GAIA fuel for the 18-month cycle is called GAIA 17x17, and for the 24-month cycle is called Advanced Fuel Management (AFM-GAIA) 17x17.

From the criticality standpoint, the main differences between these two types of GAIA fuel are the

[ ]. The enrichment limit for GAIA 17x17 is 5 wt%. The enrichment limit for AFM-GAIA 17x17 is contained in topical report ANP-10353P (Ref. 6.1.7). CEG is evaluating the GAIA fuel type up to an enrichment of 6.5 wt%. Also, the

[ ].

2.2.4 Computer Codes

The current analyses were performed using the MCNP4a, KENO5a, and CASMO4 suite of codes.

In order to align with more modern industry standards and common tools, the analyses underlying this request were performed using MCNP6.2 and CASMO5. MCNP6.2 has been previously used for multiple submittals throughout the industry (for example, ML23094A269) and CASMO5 was NRC approved in ML17236A419.

3.0 TECHNICAL EVALUATION

3.1 Analysis Method

For all proposed revised results, the guidance from NEI 12-16, "Guidance for Performing Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," and Regulatory Guide 1.240, "Fresh and Spent Fuel Pool Criticality Analyses" (Refs. 6.1.4 and 6.1.5) has been applied.

All racks are analyzed using the MCNP6.2 Monte Carlo neutron transport program and ENDF/B-VII.1 cross-section library. MCNP6.2 is a three-dimensional, general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies.

Page 5 of 54 ATTACHMENT 1 Evaluation of Proposed Change

CASMO5 with ENDF/B-VII.1 cross-section library was used for fuel depletion calculations to determine the isotopic composition of spent fuel for Region 2 criticality analysis. CASMO5 is a two-dimensional multigroup transport theory code for burnup calculations of BWR and PWR fuel assemblies. CASMO5 has the capability to model a planar cross section of an individual fuel assembly.

The approach specified in NEI 12-16 (Ref. 6.1.4) and RG 1.240 (Ref. 6.1.5), with some additional conservatisms, was used to perform criticality analysis for the NFV and SFP Region 1 and Region

2. The maximum keff (kmax(95/95)) was determined using Equation 1.

Equation 1

= + +

As can be seen from Equation 1, uncertainties were statistically combined (assuming that such uncertainties are mutually independent) while biases were summed up. Conservatively, only positive reactivity differences are added to calculate the final kmax(95/95).

3.2 MCNP6.2 Benchmarking

MCNP6.2 was benchmarked against a number of experimental standards as shown in Table 1.

Page 6 of 54 ATTACHMENT 1 Evaluation of Proposed Change

Table 1 Benchmarking Experiments, Numbers of Modeled Experiments, and Case Numbers Benchmarking Number of Experiment Modeled Case Number Experiments

((

))

((

))

The abovementioned cases were used to validate MCNP6.2 for various fuel and pool parameters.

((

)) were specifically used to validate MCNP6.2 for criticality calculations utilizing burnup credit. Table 2 provides a summary of the area of applicability for Light Water Reactor (LWR) type critical experiments in this LAR.

The input validation covers the ranges/parameters given in Table 2.

Page 7 of 54 ATTACHMENT 1 Evaluation of Proposed Change

Table 2 LWR Type Critical Experiments Area of Applicability Parameter Range of Values Fuel Type UO2, PuO2-UO2 Rods

Aluminum, Clad Material Stainless Steel, Zircalloy Moderator H2O Enrichment (U-235 wt%) 0.71 - 10 Pitch (cm) 0.62 - 2.64 Moderator/Fuel Volume Ratio 0.148 - 10.754

Moderator/Fuel Atom Ratio 17.4 - 629

Soluble Boron (ppm) 0 - 3389 Aluminum, Boral, Boraflex, Cd, Neutron Absorber Plates Hf, Stainless Steel, Borated Steel, Cu-Cd, Zircalloy

B4C, Absorber Rods Ag-In-Cd, UO2-Gd2O3 Pb, Reflector Walls Depleted U, Stainless Steel

The validation methodology was based on recommendations contained in NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculation Methodology" (Ref. 6.2.1). The normal distribution was used in this analysis for the calculated keff values. If the data was normally distributed, then a technique such as a one-sided tolerance limit was used to determine the upper safety limit (USL). If the data was not normally distributed, then a non-parametric analysis method was used to determine the USL. Shapiro-Wilk test was used for cases with fewer than 50 samples to test the hypothesis that the calculated keff values are normally distributed about the mean keff, and otherwise Chi-square test was used.

Table 3 summarizes the parameters, their correlations with keff, and their distributions (if normal).

Table 4 summarizes the biases and bias uncertainties for the validated MCNP6.2, ((

Page 8 of 54 ATTACHMENT 1 Evaluation of Proposed Change

))

Table 3 Summarized Parameters, Correlations and Distributions Correlation Significant Normal Correlated Parameter, x No. of Exp. Coefficient, Correlation Distribution R2

((

Page 9 of 54 ATTACHMENT 1 Evaluation of Proposed Change

Table 3 (continued)

Summarized Parameters, Correlations and Distributions

))

Page 10 of 54 ATTACHMENT 1 Evaluation of Proposed Change

Table 4 Summarized Biases and Bias Uncertainties Set Subsets Bias Bias Uncertainty

((

)) 0.0045

((

)) -0.0026 ((

))

The bias of 0.0026 and bias uncertainty of 0.0045 were conservatively used in both NFV and SFP criticality analyses. (( )).

3.3 New Fuel Vault Criticality Analysis

The NFV criticality analysis demonstrates that the storage rack maximum keff (kmax(95/95))

remains below regulatory limits for all normal, abnormal, and credible accident conditions with biases and uncertainties included, when the storage racks are fully loaded with fuel with the maximum fuel assembly anticipated reactivity.

Page 11 of 54 ATTACHMENT 1 Evaluation of Proposed Change

The GAIA 17x17 fuel with 6.5 wt% enrichment was assumed as the base case fuel. The nominal values for GAIA 17x17 fuel were used to build the MCNP model of the base case fuel assembly.

The M5 cladding was conservatively modeled as pure Zirconium (Zr). Fuel blankets were conservatively replaced by fuel with an enrichment equal to the fuel central zone. Spacer grids were replaced by water. The dashpot at the bottom of the guide tube was modeled with the same diameter as that of the top part of the guide tube. Conservatively, no fuel burnable absorber rods were credited for base case calculations. It should also be noted that the area of a guide tube was modeled 0.66% less than the actual. Since guide/instrumentation tubes area is a small portion of the assembly cross section area, the change in the amount of displaced water is negligible, and the reactivity effect of this deviation is insignificant. Figure 1 shows the radial cross section view of MCNP model of the base case fuel.

Figure 1 Radial Cross Section View of MCNP Model of the Base Case Fuel -

Active Fuel Region (Green is the rack box, light blue is water, purple is fuel)

Mostly, the nominal values were used to build the MCNP model of the base case NFV racks. The NFV dimensions were modeled less than the nominal. Conservatively, most of the NFV structural materials and the section of the fuel box above the fuel assembly were not modeled. The MCNP model on the NFV racks includes the NFV cavity with three 22x2 new fuel cell arrays, filled with water, radially surrounded by concrete. Under the racks, steel was modeled, and above the racks, water (with the corresponding water density of the remaining part of the model) and steel were modeled. The radial and axial cross sections of the MCNP model are shown in Figures 2 and 3, respectively.

Page 12 of 54 ATTACHMENT 1 Evaluation of Proposed Change

Figure 2 Radial Cross Section View of MCNP Model of the NFV (Orange is concrete, and light blue and white are water)

Figure 3 Axial Cross Section View of MCNP Model of the NFV (Orange is concrete, light blue and white are water, green is steel)

To determine the bounding water densities, MCNP calculations were performed with the following water densities in g/cm3:

1, 0.9, 0.8, 0.7, 0.6, 0.5, 0.4, 0.3, 0.2, 0.18, 0.16, 0.14, 0.12, 0.1, 0.09, 0.08, 0.07, 0.06, 0.05, 0.045, 0.04, 0.035, 0.03, 0.025, 0.02, 0.015, 0.01.

Page 13 of 54 ATTACHMENT 1 Evaluation of Proposed Change

Then, the MCNP calculations to determine the applicable biases and uncertainties were performed for the full water density and the case with the bounding optimum-moderator density (which was 0.04 g/cm3). Figure 4 shows how the keff was changed as a function of the water density.

Figure 4 keff as a Function of Water Density

Consistent with NEI 12-16, the following biases were considered:

x Criticality code validation bias.

x Base case fuel assembly bias (The keff difference between GAIA 17x17 and AFM-GAIA 17x17) x Eccentric positioning bias (Six scenarios were modeled, as shown in Figures 5-10; only the four assemblies at the center of the middle array are shown in the figures.)

x NFV base and perimeter bias (on how NFV cavity was modeled compared to the actual) x Temperature reactivity effect (WRHYDOXDWHWKHHIIHFWRI6 )

Page 14 of 54 ATTACHMENT 1 Evaluation of Proposed Change

Figure 5 Radial Cross Section View of MCNP Model of Eccentric Positioning; fuel assemblies were inward the center of the middle array.

Figure 6 Radial Cross Section View of MCNP Model of Eccentric Positioning; fuel assemblies were outward the center of the middle array.

Page 15 of 54 ATTACHMENT 1 Evaluation of Proposed Change

Figure 7 Radial Cross Section View of MCNP Model of Eccentric Positioning; fuel assemblies were inward X-direction.

Figure 8 Radial Cross Section View of MCNP Model of Eccentric Positioning; fuel assemblies were outward X-direction.

Page 16 of 54 ATTACHMENT 1 Evaluation of Proposed Change

Figure 9 Radial Cross Section View of MCNP Model of Eccentric Positioning; fuel assemblies were inward Y-direction.

Figure 10 Radial Cross Section View of MCNP Model of Eccentric Positioning; fuel assemblies were outward Y-direction.

The temperature of 4 °C was considered for base case calculations. To determine the effect of temperature on reactivity (temperature reactivity effect), a sensitivity study was performed for temperatures of 20, 80 and 124 °C. These temperatures were selected since MCNP includes their thermal treatment (LWTR, 6 DWWKHVHRUQHDUE\\WHPSHUDWXUHV  °C is also recommended in NEI 12-16 for the SFP (Ref. 6.1.4). The water densities for full-moderator cases were adjusted accordingly. For the optimum moderator density, the maximum reactivity was at 124 °C. At this temperature, three other water densities of 0.03, 0.06 and 0.15 g/cm3 were evaluated. It is shown

Page 17 of 54 ATTACHMENT 1 Evaluation of Proposed Change

that the maximum reactivity is at 124 °C and water density of 0.04 g/cm3. The reactivity difference between the bounding case and the base case, if positive, was added as a bias when calculating kmax(95/95).

The evaluated uncertainties are provided below. This list covers applicable uncertainties provided in NEI 12-16 (Ref. 6.1.4).

x Fuel manufacturing tolerances (95/95) o Fuel pitch o Fuel pellet diameter o Fuel clad ID o Fuel clad OD o Guide tube ID o Guide tube OD o Fuel enrichment o Fuel density x Rack manufacturing tolerances (95/95) o Box ID o Box wall thickness o Cell pitch o Distance between two arrays x Criticality code validation uncertainty (the bounding values from the benchmarking report).

x Monte Carlo calculational uncertainty (2 sigma)

With an exception, the upper and lower bounds of fuel and rack manufacturing tolerances were analyzed in MCNP. The difference between the larger keff of the upper and lower bounds cases and the base case keff was used as an uncertainty, if larger than 0. Otherwise, the uncertainty of 0 was used. The exception was fuel enrichment which only the upper bound was analyzed, considering lower enrichment fuel will result in a lowered reactivity.

Using Equation 1, the kmax(95/95) values for both full-water density and optimum-moderator water density were calculated. The results are summarized in Table 5. It is demonstrated that the kmax (95/95) values meet the regulatory limits.

Page 18 of 54 ATTACHMENT 1 Evaluation of Proposed Change

Table 5 kmax(95/95) Calculation Full-moderator density Base Case keff 0.93297 Biases Max. Difference Criticality code validation bias 0.00260 Base case bias 0.00102 Eccentric positioning bias 0.00405 NFV base and perimeter modeling bias 0 Temperature reactivity effect 0 Total bias 0.00767 Uncertainties Fuel manufacturing tolerances (95/95) 0.00487 Rack manufacturing tolerances (95/95) 0.00286 Criticality code validation uncertainty 0.00450 Monte Carlo calculational uncertainty (2 sigma) 0.00036 Statistical combination of uncertainties 0.00723

kmax(95/95) 0.9479 Regulatory Limit 0.95 Optimum-moderation water density Base Case keff 0.76364 Biases Max. Difference Criticality code validation bias 0.00260 Base case bias 0.00191 Eccentric positioning bias 0.00105 NFV base and perimeter modeling bias 0.00851 Temperature reactivity effect 0.03133 Total bias 0.04540 Uncertainties Fuel manufacturing tolerances (95/95) 0.00286 Rack manufacturing tolerances (95/95) 0.00921 Criticality code validation uncertainty 0.00450 Monte Carlo calculational uncertainty (2 sigma) 0.00032 Statistical combination of uncertainties 0.01065

kmax(95/95) 0.8197 Regulatory Limit 0.98

All calculations stated above were performed at a bounding enrichment of 6.5 wt%, with no credit for rods with gadolinium oxide. Additional calculations were done to determine how kmax(95/95) varies as a function of the fuel enrichment, from 5 wt% to 6.5 wt%, as shown in Figure 11. ((

Page 19 of 54 ATTACHMENT 1 Evaluation of Proposed Change

)). The results of these calculations are also shown in Figure 11. This figure shows the actual margin to the limit is more than 0.02.

Figure 11 Reactivity Effect of U-235 Enrichment and Gd Rods - Full Moderator Density

3.4 Region 1 Criticality Analysis

The SFP Region 1 criticality analysis demonstrates that the storage rack maximum keff (kmax(95/95)) remains below regulatory limits for all normal, abnormal, and credible accident conditions with biases and uncertainties included, when the storage racks are fully loaded with fuel with the maximum fuel assembly anticipated reactivity.

MCNP calculations were performed for SFP filled with fresh water (Unborated water) and SFP filled with 500 ppm soluble boron (borated water). Consistent with NEI 12-16, an additional 50 ppm of soluble boron needs to be reserved to offset the reactivity impact of the fuel assembly grids. This 50 ppm is also sufficient to offset the change in reactivity effect of tolerances under borated conditions. Thus, 550 ppm soluble boron was required for criticality analysis of the SFP Region 1.

The GAIA 17x17 fuel with 6.5 wt% enrichment was assumed as the base case fuel. The nominal values for GAIA 17x17 fuel were used to build the MCNP model of the base case fuel assembly.

The reflective boundary conditions were conservatively used both radially and axially. Axially, only the active part of fuel was conservatively modeled. The M5 cladding was conservatively

Page 20 of 54 ATTACHMENT 1 Evaluation of Proposed Change

modeled as pure Zr. Fuel blankets were conservatively replaced by fuel with an enrichment equal to the fuel central zone. Spacer grids were replaced by water. The pellet nominal density was conservatively used instead of the nominal stack density. The dashpot at the bottom of the guide tube was modeled with the same diameter as that of the top part of the guide tube. Conservatively, no fuel burnable absorber rods were credited for base case calculations. It should also be noted that the area of a guide tube was modeled 0.66% less than the actual. Since guide/instrumentation tubes area is a small portion of the assembly cross section area, the change in the amount of displaced water is negligible, and the reactivity effect of this deviation is insignificant. Figure 12 shows the radial cross section view of MCNP model of the base case fuel.

Figure 12 Radial Cross Section View of MCNP Model of Base Case Fuel Assembly

The nominal values were used to build the MCNP model of the base case Region 1 racks. The base case model included a 2x2 array Region 1 cells. Axially, only the active region of the fuel assembly was modeled. Reflective boundary conditions were used on all 6 sides, making the models infinite.

Consistent with NEI 12-16, the following biases were considered:

x Criticality code validation bias (the bounding values from the benchmarking report) x Base Case fuel assembly (the keff difference between GAIA 17x17 and AFM-GAIA 17x17) x Eccentric positioning bias (four scenarios were modeled as shown in Figures 13-16).

x Moderator temperature bias.

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Figure 13 Radial Cross Section View of MCNP Model of an Infinite 2x2 Array Eccentric Positioning, Fuel Assemblies Toward the Center, Reflective Boundary Conditions

Figure 14 Radial Cross Section View of MCNP Model of an Infinite 2x2 Array Eccentric Positioning, Fuel Assemblies Toward a Corner, Periodic Boundary Conditions

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Figure 15 Radial Cross Section View of MCNP Model of an Infinite 8x8 Array Eccentric Positioning, Fuel Assemblies Toward the Center, Reflective Boundary Conditions

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Figure 16 Radial Cross Section View of MCNP Model of an Infinite 8x8 Array Eccentric Positioning, Fuel Assemblies Toward the Corners, Reflective Boundary Conditions

The 4 °C moderator temperature was conservatively used for the base case model (base case fuel and rack). By changing the water density (up to the water density at 124 °C and 20% void) and temperature in MCNP, the effect of the moderator (water) density on keff was evaluated. The evaluated temperature range covers the range for both normal and accident conditions. ((



)) The difference between the maximum keff of the moderator-

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temperature cases and the base case keff was used as the moderator-temperature bias, if larger than 0. Otherwise, the bias of 0 was used.

The evaluated uncertainties are provided below. This list covers applicable uncertainties provided in NEI 12-16 (Ref. 6.1.4).

x Fuel manufacturing tolerances (95/95) o Fuel pitch o Fuel pellet diameter o Fuel clad ID o Fuel clad OD o Guide tube ID o Guide tube OD o Fuel enrichment o Fuel density x Rack manufacturing tolerances (95/95) o Box ID (change in flux trap) o Box wall thickness o Poison width o Poison thickness (resulting in a boron areal density less than the minimum) o Cell pitch o Poison (Boral) coupon B-10 degradation x Criticality code validation uncertainty (the bounding values from the benchmarking report) x Monte Carlo calculational uncertainty (2 sigma)

With few exceptions, the upper and lower bounds of fuel and rack manufacturing tolerances were analyzed in MCNP. The difference between the larger keff of the upper and lower bounds cases and the base case keff was used as an uncertainty, if larger than 0. Otherwise, the uncertainty of 0 was used. The exceptions were fuel enrichment which only the upper bound was analyzed, and poison width which only lower bound was analyzed, considering lower enrichment fuel and wider poison will result in a lowered reactivity.

Using Equation 1, the kmax(95/95) values for borated and unborated water are calculated and summarized in Table 6. It is demonstrated that the kmax(95/95) values for both borated and unborated water meet the regulatory limits.

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Table 6 kmax(95/95) Values for Borated and Unborated Water Borated water Base case k eff 0.91624 Biases Criticality code validation bias 0.00260 Base case bias 0.00232 Moderator temperature bias 0 Eccentric positioning bias 0 Total bias 0.00392 Uncertainties Fuel manufacturing tolerances 0.00609 Rack manufacturing tolerances 0.00810 Criticality code validation uncertainty 0.00450 Monte Carlo calculational uncertainty (2 sigma) 0.00038 Statistical combination of uncertainties 0.01109

kmax(95/95) (Note 1) 0.9323 Regulatory Limit 0.95 Unborated water Base case k eff 0.96242 Biases Criticality code validation bias 0.00260 Base case bias 0.00080 Moderator temperature bias 0 Eccentric positioning bias 0 Total bias 0.00340 Uncertainties Fuel manufacturing tolerances 0.00629 Rack manufacturing tolerances 0.00711 Criticality code validation uncertainty 0.00450 Monte Carlo calculational uncertainty (2 sigma) 0.00038 Statistical combination of uncertainties 0.01051

kmax(95/95) (Note 1) 0.9764 Regulatory Limit 1

The evaluated abnormal and accident conditions were temperature beyond the normal operating range, mislocated fuel assembly, assembly misload, and dropped fuel assembly. It is shown that these abnormal and accident conditions are either not credible or do not result in an increased reactivity.

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Region 1 storage racks only contain uniform loading configuration; thus the bounding interface configuration could only have fresh fuel assemblies adjacent to each other as the same as base case. In addition, there is a small gap between neighboring Region 1 style racks. Hence the distance between two Region 1 racks across any interface is more than the flux trap between two Region 1 cells analyzed in the base model. The reactivity of the Region 1 to Region 1 interface is bounded by the base cases.

3.5 Region 2 Criticality Analysis

The SFP Region 2 criticality analysis demonstrate that the storage rack maximum keff (kmax(95/95))

remains below regulatory limits for all normal, abnormal, and credible accident conditions with biases and uncertainties included, when the storage racks are fully loaded with fuel with the maximum fuel assembly anticipated reactivity.

The AFM-GAIA 17x17 fuel with enrichment up to 6.5 wt% was assumed as the base case fuel, since AFM-GAIA fuel bounds other GAIA fuel assembly types stored in Braidwood and Byron SFPs. The nominal values for AFM-GAIA 17x17 fuel were used to build the MCNP model of the base case fuel assembly. The base case model included a 2x2 array Region 2 cells. Axially, only the active region of the fuel assembly was modeled. Any neutron absorbers and structural components above and below the active fuel region were conservatively replaced by unborated water. Radially, periodic boundary was considered for all four sides to make the models infinite.

The M5 cladding was conservatively modeled as pure Zr. Fuel blankets were conservatively replaced by fuel with an enrichment equal to the fuel central zone. Spacer grids were replaced by water. The pellet nominal density was conservatively used instead of the nominal stack density.

The dashpot at the bottom of the guide tube was modeled with the same diameter as that of the top part of the guide tube. In the analysis, it is assumed the ((

)). The area of a guide tube was modeled 0.66% less than the actual. Since guide/instrumentation tubes area is a small portion of the assembly cross section area, the change in the amount of displaced water is negligible, and the reactivity effect of this deviation is insignificant. Figure 17 shows the radial cross section view of MCNP model of the base case fuel.

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Figure 17 Radial Cross Section View of MCNP Model of Base Case Fuel Assembly in Region 2 Racks

The nominal values were used to build the MCNP model of the base case Region 2 racks. The base case model included a 2x2 array of Region 2 cells. Axially, only the active region of the fuel assembly was modeled. Water reflector was modelled above and below the active fuel region, and reflective boundary conditions were used above and below water reflector. Radially, periodic boundary was considered for all four sides, making the models infinite. This was conservative since the distance between any Region 2 racks was neglected.

Various scenarios of eccentric fuel positioning were modeled as shown in Figures 18-20. It was concluded that all assemblies centered in their fuel storage cell of a 2x2 array was bounding.

Therefore, the bounding eccentric position of the fuel assemblies is included in the base case calculations.

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Figure 18 Radial Cross Section View of MCNP Model of an infinite 2x2 array Eccentric Positioning, Fuel Assemblies Were Inward Each Other

Figure 19 Radial Cross Section View of MCNP Model of an infinite 2x2 array Eccentric Positioning, Fuel Assemblies Were Outward to the Same Corner of the Cell

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Figure 20 Radial Cross Section View of MCNP Model of an infinite 8x8 array Eccentric Positioning, Fuel Assemblies Were Inward Each Other

The 4 °C moderator temperature was used for the base case model (base case fuel and base case rack). By changing the water density and its associated temperature card in MCNP, the effect of the moderator density on keff was evaluated. The evaluated temperature range covers the range for both normal and accident conditions. ((



)) It was concluded that the minimum temperature used in the base case calculations was bounding.

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CASMO5 was used for depletion calculation to calculate the isotopic composition of the spent fuel materials so that they could be used as input data of the criticality calculations of spent fuel.

Isotopic compositions were calculated with CASMO5 for a range of enrichments ((

)). Depletion calculations were performed with conservative operating conditions: highest fuel temperature, moderator temperature and soluble boron concentrations during in-core operation. Gadolinia burnable absorbers were also conservatively neglected. Linear interpolation was used to determine the isotopic composition for the intermediate burnups. The assembly average isotopic compositions were applied to all fuel rods of spent fuel assemblies in the MCNP models.

For the determination of the maximum keff (kmax(95/95)), the applicable biases were provided below and summed up in Equation 1. This list covers applicable biases provided in NEI 12-16 (Ref. 6.1.4).

x Criticality code validation bias (the bounding values from the benchmarking report).

x Axial burnup profile bias. Base case calculations were performed with axial burnup profile from NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses" (Ref. 6.2.6). The results were compared to the axially uniform profile, which assumes the same burnup along the entire axial length. The reactivity difference between the cases using uniform profile and the base case was considered as the bias.

x Fuel depletion related geometry change bias. The positive (if any) reactivity effects of fuel grid growth as well as fuel rod growth and cladding creep were added together as bias.

x Fission gas release bias. The reactivity difference between the cases with the fraction of fission gases removed and the base case was considered as bias.

The applicable uncertainties were provided below and statistically combined in Equation 1 since they were mutually independent. This list covers applicable uncertainties provided in NEI 12-16 (Ref. 6.1.4).

x Fuel manufacturing tolerances. The upper and lower bounds of fuel manufacturing tolerances were analyzed. The difference between the larger keff of the upper and lower bounds cases and the base case keff was used as an uncertainty, if larger than 0.

Otherwise, the uncertainty of 0 was used. The exceptions were fuel enrichment and density which only the upper bound was analyzed, and poison width which only lower bound was analyzed, considering lower enrichment and density fuel will result in a lowered reactivity.

o Fuel rod pitch o Fuel pellet diameter o Fuel clad ID o Fuel clad OD o Guide/Instrument tube thickness o Fuel enrichment o Fuel density

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x Rack manufacturing tolerances. The upper and lower bounds of rack manufacturing tolerances were analyzed. The difference between the larger keff of the upper and lower bounds cases and the base case keff was used as an uncertainty, if larger than 0.

Otherwise, the uncertainty of 0 was used. The exceptions were poison width and thickness which only lower bound was analyzed, considering larger poison thickness and width will result in a lowered reactivity.

o Box ID o Box wall thickness o Poison width o Poison thickness o Cell pitch o Poison (Boral) coupon B-10 degradation x Criticality code validation uncertainty (the bounding values from the benchmarking report) x Monte Carlo calculational uncertainty (2 sigma) x Fuel depletion related geometry change uncertainty (95/95 uncertainty of the bias) x Failed cladding uncertainty. This uncertainty was not listed in NEI 12-16. It was assumed unborated water filled the fuel gap to represent fuel rods with failed cladding.

x Depletion uncertainty. The uncertainty was determined by multiplying 5% with the reactivity difference (with 95/95 uncertainty) between the base case with spent fuel and a corresponding case with fresh fuel at the same fuel enrichment. Recent studies performed by EPRI (Ref. 6.2.2) indicated that the 5% depletion uncertainty was conservative in SFP criticality calculations.

x Burnup uncertainty. The uncertainty was determined by the reactivity difference (with 95/95 uncertainty) between the base case with spent fuel and a corresponding case with spent fuel with the fuel burnup 5% lower at the same enrichment.

x Fission gas release uncertainty (95/95 uncertainty of the bias)

The kmax(95/95) values were calculated based on the calculated keff, the applicable bias, and the applicable uncertainties using Equation 1. Two base case calculations were performed, one at lower bound burnup and the other at upper bound burnup (with burnup increment of 5 GWd/MTU),

for each selected enrichment. Linear interpolation between the two designated burnup values was used to determine the specific burnup for the target kmax(95/95) value of 0.995 (See Table 7). The final burnups versus enrichments were fitted to a curve by a third-order polynomial and this curve was adjusted conservatively such that all calculated burnups were on or below the curve. Table 7 shows the base case calculations used to generate the loading curve and the results of the polynomial burnup values. The resulting equation for the curve fits is summarized in Table 8, and graphically shown in Figure 21. The loading curve was validated by confirmatory calculations with all enrichments shown in Table 7 for the conditions with unborated water, and the case with 500 ppm borated water. It was confirmed the largest kmax (95/95) value is 0.99421 with fresh water and 0.94781 with 500 ppm borated water, thus below the regulatory limit. Consistent with NEI 12-16, under borated conditions, an additional 50 ppm of soluble boron needs to be reserved to offset the reactivity impact of the fuel assembly grids, and this 50 ppm is also sufficient to offset the change in reactivity effect of tolerances. Thus, 550 ppm soluble boron was required under normal conditions for the SFP Region 2.

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Table 7 kmax(95/95) Calculation for Region 2 Enrichment (wt%) 2 2.5 3 3.5 4 Lower Bound Burnup (GWd/MTU) 0 5 15 20 25 Upper Bound Burnup (GWd/MTU) 5 10 20 25 30 keff at Lower Bound Burnup 0.96558 0.98109 0.98023 0.96825 0.97593 keff at Upper Bound Burnup 0.92306 0.94156 0.93100 0.94292 0.95222 Biases Criticality Code Validation Bias 0.00260 0.00260 0.00260 0.00260 0.00260 Fuel Depletion Geometry Change Bias 0.00800 0.00800 0.00800 0.00800 0.00800 Fission Gas Release Bias 0.00024 0.00075 0.00110 0.00126 0.00105 Axial Burnup Profile Bias 0.00224 0.00310 0.00000 0.00000 0.00000 Total Bias 0.01308 0.01445 0.01170 0.01186 0.01165 Uncertainties MCNP calculational uncertainty 0.00064 0.00066 0.00066 0.00068 0.00072 Fuel Manufacturing Tolerances Uncertainty 0.00963 0.00963 0.00963 0.00963 0.00963 Rack Manufacturing Tolerances Uncertainty 0.00729 0.00729 0.00729 0.00729 0.00729 Criticality Code Validation Uncertainty 0.00450 0.00450 0.00450 0.00450 0.00450 Fuel Depletion Geometry Change Unc. 0.00138 0.00138 0.00138 0.00138 0.00138 Fission Gas Release Uncertainty 0.00089 0.00095 0.00098 0.00098 0.00103 Failed Cladding Uncertainty 0.00436 0.00436 0.00436 0.00436 0.00436 Depletion Uncertainty 0.00217 0.00441 0.00746 0.00876 0.00992 Burnup Uncertainty 0.00255 0.00479 0.00601 0.00618 0.00703 Statistical Combination of Uncertainties 0.01412 0.01519 0.01674 0.01742 0.01834

Total Biases and Uncertainties 0.02720 0.02964 0.02844 0.02928 0.02999 Targeted kmax (95/95) 0.99500 0.99500 0.99500 0.99500 0.99500 Targeted keff 0.96780 0.96536 0.96656 0.96572 0.96501 Interpolated Burnup (GWd/MTU) 0.00 6.99 16.39 20.50 27.30 Polynomial Burnup (GWd/MTU) 4.08 11.57 18.42 24.71 30.49

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Table 7 (Continued) kmax(95/95) Calculation for Region 2 Enrichment (wt%) 4.5 5 5.5 6 6.5 Lower Bound Burnup (GWd/MTU) 30 35 40 45 50 Upper Bound Burnup (GWd/MTU) 35 40 45 50 55 keff at Lower Bound Burnup 0.98118 0.98023 0.97160 0.97333 0.97413 keff at Upper Bound Burnup 0.95401 0.94643 0.95156 0.95391 0.95607 Biases Criticality Code Validation Bias 0.00260 0.00260 0.00260 0.00260 0.00260 Fuel Depletion Geometry Change Bias 0.00800 0.00800 0.00800 0.00800 0.00800 Fission Gas Release Bias 0.00030 0.00175 0.00224 0.00189 0.00212 Axial Burnup Profile Bias 0.00000 0.00000 0.00000 0.00000 0.00000 Total Bias 0.01090 0.01235 0.01284 0.01249 0.01272 Uncertainties MCNP calculational uncertainty 0.00068 0.00072 0.00076 0.00074 0.00072 Fuel Manufacturing Tolerances Uncertainty 0.00963 0.00963 0.00963 0.00963 0.00963 Rack Manufacturing Tolerances Uncertainty 0.00729 0.00729 0.00729 0.00729 0.00729 Criticality Code Validation Uncertainty 0.00450 0.00450 0.00450 0.00450 0.00450 Fuel Depletion Geometry Change Unc. 0.00138 0.00138 0.00138 0.00138 0.00138 Fission Gas Release Uncertainty 0.00096 0.00102 0.00103 0.00098 0.00098 Failed Cladding Uncertainty 0.00436 0.00436 0.00436 0.00436 0.00436 Depletion Uncertainty 0.01116 0.01267 0.01342 0.01413 0.01479 Burnup Uncertainty 0.00793 0.00932 0.01072 0.01041 0.01128 Statistical Combination of Uncertainties 0.01938 0.02088 0.02199 0.02228 0.02312

Total Biases and Uncertainties 0.03028 0.03323 0.03483 0.03477 0.03584 Targeted kmax (95/95) 0.99500 0.99500 0.99500 0.99500 0.99500 Targeted keff 0.96472 0.96177 0.96017 0.96023 0.95916 Interpolated Burnup (GWd/MTU) 33.03 37.73 42.85 48.37 54.14 Polynomial Burnup (GWd/MTU) 35.84 40.83 45.51 49.96 54.25

Table 8 Loading Curve for Region 2 Loading Minimum Required Fuel Assembly Burnup (GWd/MTU) as a Function of Configurations Initial Enrichment (wt% U-235)

Uniform Loading for f(x) = 0.0894x3 - 1.9403x2 + 22.344x - 33.56 (Note 1)

Region 2

Note 1. x is defined as initial enrichment, while f(x) is minimum required fuel burnup.

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Figure 21 Loading Curve for Region 2

The normal operation scenarios were discussed as follows:

x Fuel movement and inspection is controlled by procedures onsite to make sure a fuel assembly is always moved above the spent fuel racks and isolated from any other fuel assemblies, Therefore, such operation in the spent fuel pool is bounded by a single fresh fuel assembly in water for which the reactivity is below the regulatory limit.

x For spent fuel assembly, the most reactive fuel region is at the top of the fuel, hence the axial segments of fuel assemblies dominating in reactivity were perfectly aligned. This is also applicable to fresh fuel assembly since the most active fuel region is in the middle of the fuel. Therefore, moving fuel in and out of the rack cells would make the most reactive fuel region no longer aligned with other surrounding fuel assemblies thus does not increase reactivity.

x In the case that any fuel assembly needs fuel reconstitution, the activity will be performed with the assembly isolated from any other fuel assembly. Further evaluation will be performed for the reconstituted fuel assemblies separately. This is directed by Constellation Procedure NF-AP-309, "PWR Special Nuclear Material and Core Component Move Sheet Development."

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x ((

))

Table 9

((

))

Note 1: The applicable Total Biases and Uncertainties determined in Table 7 is used to determine kmax (95/95).

The reactivity effects of various abnormal and accident scenarios were evaluated.

x Under accident conditions (loss of cooling) the SFP water temperature could be elevated beyond the normal operating range. The range of temperatures for both normal and accident conditions were evaluated, and it was determined that the 4 °C moderator temperature used for the base case was bounding.

x ((

)) As a bounding approach, the mislocation of a single fresh fuel assembly with an enrichment of 6.5 wt% in a rack corner facing two adjacent fuel assemblies was considered. All exterior storage rack walls and the attached

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Boral panels at the rack corner were credited (see Figure 22). Fuel assemblies with various enrichments and corresponding burnups calculated using the polynomial functions of loading curve were evaluated with fresh and borated water. The minimum soluble boron concentration was determined to ensure that the kmax(95/95) value does not exceed the regulatory limit of 0.95.

x The misplacement of a fresh fuel assembly into a Region 2 storage cell intended for spent fuel, without soluble poison, could result in reactivity exceeding the regulatory limit. As a bounding approach, the misload of a single fresh fuel assembly with an enrichment of 6.5 wt% was considered in a storage cell close to the rack center that provides the largest positive reactivity increase. The calculational models consist of a large array of assemblies with periodic boundary conditions on all four sides, which effectively represents a multiple misload. Fuel assemblies with various enrichments and corresponding burnups calculated using the polynomial functions of loading curve are evaluated with fresh and borated water. The minimum soluble boron concentration was determined to ensure that a kmax(95/95) value does not exceed the regulatory limit of 0.95.

x Multiple fuel assemblies misload could occur because of planning or process error. In this case, administrative control and process check can be used to ensure that fresh fuel assemblies were not selected when spent fuel assemblies were intended to be loaded, thus the need to assume a multiple misload of fresh fuel was eliminated. As a bounding approach, 16 spent fuel assemblies irradiated for a single cycle with 6.5 wt% permissible enrichment were assumed to be accidentally loaded into storage cells qualified for spent fuel in the Region 2 racks. (( )) was selected as the burnup of misloaded fuel assemblies since ((

)). The evaluations were performed using the similar calculational models as the single misload accident, but 16 misloaded assemblies were considered in storage cells of the rack center to provide the largest positive reactivity increase (See Figure 23). Fuel assemblies with various enrichments and corresponding burnups calculated using the polynomial functions of loading curve are evaluated with fresh and borated water. The minimum soluble boron concentration was determined to ensure that a kmax(95/95) value does not exceed the regulatory limit of 0.95.

x For the case in which a fuel assembly is assumed to be dropped horizontally on top of the Region 2 storage racks, the fuel assembly will come to rest on top of the rack. The distance between the top of the rack and the active region is more than 12 inches, which is sufficient to preclude neutron coupling and consequently an increase in reactivity. Vertically dropping an assembly into a location occupied by another assembly would at most cause a small compression of the stored assembly, the distance between the active fuel regions of both assemblies will be more than sufficient to ensure no neutron interaction between the two assemblies. In addition, a vertical drop of fuel assembly into an empty cell may result in a small deformation of the baseplate and potentially cause reactivity increase.

However, the reactivity increase would be bounded by the misload accident discussed above.

x In the event of seismic activity, storage racks may slide and come closer to each other. In the worst-case scenario, two racks may touch each other at the baseplate, reducing the physical separation of the fuel assemblies along the rack interface, but still maintaining a minimum water gap width. The worst-case scenario is when the water gap width between

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all racks is as low as allowed by the baseplate. This accident condition is bounded by the base cases which consider all the racks at their closest approach, i.e., a laterally infinite 2x2 arrays so that the gap between the racks is neglected. Consequently, there will be no positive effect on reactivity as a result of rack movement.

The bounding results for mislocated fuel assembly, single assembly misload, and multiple assembly misload accidents are summarized in Table 10. Overall, under all credible abnormal and accident conditions, the minimum soluble boron concentration of 2000 ppm is sufficient to ensure that kmax(95/95) of the racks does not exceed the regulatory limit, with consideration of additional 50 ppm of soluble boron reserved to offset the reactivity impact of the fuel assembly grids and the change in reactivity effect of tolerances under borated conditions.

Table 10 Results of Accident Cases Soluble boron kmax Description Concentration (ppm) keff Uncertainty (95/95)

(Note 1)

Mislocated Fuel Assembly Cases, 6.5 wt%, 54.24 GWd/MTU Cell centered 0 0.96815 0.00038 1.00399 Toward the mislocated fuel 0 0.97520 0.00039 1.01104 assembly Cell centered 1000 0.87210 0.00035 0.90794 Toward the mislocated fuel 1000 0.87629 0.00037 0.91213 assembly Required Soluble Boron Concentration (ppm) 617.1 Single Assembly Misload Cases, 6.5 wt%, 54.24 GWd/MTU Cell centered 0 0.98597 0.00036 1.02181 Rack centered 0 0.99669 0.00036 1.03253 Cell centered 1000 0.90003 0.00035 0.93587 Rack centered 1000 0.91077 0.00038 0.94661 Required Soluble Boron Concentration (ppm) 960.5 Multiple Assembly Misload Cases, 6.5 wt%, 54.24 GWd/MTU Cell centered 0 1.06156 0.00037 1.09740 Rack centered 0 1.07049 0.00038 1.10633 Cell centered 1000 0.97238 0.00037 1.00822 Rack centered 1000 0.98146 0.00037 1.01730 Cell centered 2000 0.90163 0.00037 0.93747 Rack centered 2000 0.91045 0.00038 0.94629 Required Soluble Boron Concentration (ppm) 1912.2 Note 1: The applicable Total Biases and Uncertainties determined in Table 7 is used to determine kmax (95/95).

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Figure 22 Radial Cross Section View of MCNP Model for Mislocated Fuel Assembly

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Figure 23 Radial Cross Section View of MCNP Model for Multiple Assembly Misload

Interface conditions related with Region 2 storage racks were evaluated and qualified.

x Region 2 storage racks contain only uniform loading configuration of spent fuel assemblies. All region 2 racks contain neutron absorber panels on the exterior surfaces facing adjacent racks as well as small gaps between neighboring Region 2 racks, which were conservatively neglected in the base cases, Therefore, the reactivity of the any Region 2 to Region 2 rack interface was bounded by the base cases.

x For the Region 1 to Region 2 interface, each fresh fuel assembly in Region 1 rack at the interface faces a spent fuel assembly in Region 2 rack across the interface. This configuration is not bounded by either Region 1 or Region 2 base cases. Calculations were performed for the Region 1 to Region 2 interface with considerations of ((

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)). Fresh fuel assemblies in Region 1 rack were modelled with enrichment of 6.5 wt%. Spent fuel assemblies in Region 2 were evaluated with various enrichments at corresponding required burnup from loading curve. The results of the calculations summarized in Table 11 confirmed the reactivity of the interface is still below the regulatory limit. ((

))

Figure 24 (( ))

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Table 11 Results of Region 1 to Region 2 Interface Cases Soluble boron kmax Description Concentration keff Uncertainty (95/95)

(ppm) (Note 1) 2.0 wt%, 4.07 GWd/MTU All Fuel Assemblies Cell Centered 0 0.95311 0.00041 0.98895 Fuel Assemblies Toward to Interface Center 0 0.95313 0.00041 0.98897 All Fuel Assemblies Cell Centered 500 0.90752 0.00038 0.94336 Fuel Assemblies Toward to Interface Center 500 0.90639 0.00040 0.94223 3.5 wt%, 24.70 GWd/MTU All Fuel Assemblies Cell Centered 0 0.95355 0.00043 0.98939 Fuel Assemblies Toward to Interface Center 0 0.95196 0.00042 0.98780 All Fuel Assemblies Cell Centered 500 0.90794 0.00039 0.94378 Fuel Assemblies Toward to Interface Center 500 0.90680 0.00041 0.94264 5.0 wt%, 40.82 GWd/MTU All Fuel Assemblies Cell Centered 0 0.95270 0.00041 0.98854 Fuel Assemblies Toward to Interface Center 0 0.95243 0.00043 0.98827 All Fuel Assemblies Cell Centered 500 0.90736 0.00042 0.94320 Fuel Assemblies Toward to Interface Center 500 0.90705 0.00040 0.94289 6.5 wt%, 54.24 GWd/MTU All Fuel Assemblies Cell Centered 0 0.95298 0.00040 0.98882 Fuel Assemblies Toward to Interface Center 0 0.95228 0.00040 0.98812 All Fuel Assemblies Cell Centered 500 0.90785 0.00039 0.94369 Fuel Assemblies Toward to Interface Center 500 0.90563 0.00041 0.94147 Note 1: The applicable Total Biases and Uncertainties determined in Table 7 is used to determine kmax (95/95).

3.6 Boron Dilution Event

The SFP boron dilution analysis validates that sufficient time is available to detect and terminate a boron dilution event before the regulatory criticality limit of 0.95 keff with a 95% probability at a 95% confidence level is exceeded in the Byron and Braidwood Spent Fuel Pool (SFP) storage racks containing Framatome GAIA fuel. The SFP storage rack criticality analyses (Sections 3.4 and 3.5) have determined that a minimum boron concentration of 550 ppm (plus an uncertainty of 25 ppm) is required. Thus, the boron dilution analysis is based on 575 ppm soluble boron.

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3.6.1 SFP Boron dilution Model

SFP Boron dilution can be modeled as a feed-and-bleed operation (constant volume and constant dilution flow rate). The time for boron dilution can be calculated based on the following equation:

TEnd = (V/Q)

• ln(C0/CEnd),

where:

C0 = the boron concentration of the SFP volume at the beginning of the dilution event (2,000 ppm );

CEnd = the endpoint boron concentration (575 ppm);

V = Volume (gallons) of the SFP (456,000 gallons);

Q = dilution flow rate (Gallons/minute);

TEnd = time to reach (minutes).

The SFP boron dilution analysis assumes thorough mixing of all the non-borated water added to the SFP with the contents of the SFP. If mixing is not adequate, it would be conceivable that a localized pocket of non-borated water could form somewhere in the SFP. This scenario is analyzed in Section 3.4 and 3.5, which show that the SFP storage rack keff will be less than 1.0 with the SFP filled with non-borated water.

To dilute the SFP water volume of 456,000 gallons from boron concentration of 2000 ppm to 575 ppm, it would take 2,842.09 minutes by assuming a dilution flow rate of 200 gpm and would conservatively require 568,000 gallons of non-borated water based on a feed-and-bleed operation.

3.6.2 SFP Instrumentation

Indication of SFP water level is available in the Main Control Room at Byron and Braidwood.

Additional instrumentation is provided to monitor the pressure and flow of the SFP cleanup system, and pressure and temperature of the SFP cooling system. The water level instrumentation alarms, high and low level are annunciated in the control room. The instrumentation which monitors radiation levels in the SFP area provides high radiation alarms locally in the SFP enclosure and in the control room.

A change of one foot in SFP level with the dry cask loading pit and the transfer canal isolated requires approximately 14,112 gallons of water. If the SFP level is raised from the low-level alarm point to the high-level alarm (7.5") a dilution of approximately 8,820 gallons could occur before an alarm would be received in the control room.

3.6.3 Administrative Control

The following administrative controls are in place to control the spent fuel pool boron concentration and water inventory.

1. Procedures are available to aid in the identification and termination of dilution events.

2. The procedures for loss of inventory (other than evaporation) specify that borated makeup source be used as makeup sources. The procedures specify that non-borated sources only be used as a last resort.

Page 43 of 54 ATTACHMENT 1 Evaluation of Proposed Change

3. In accordance with procedures, plant personnel perform rounds in the SFP area in the Fuel Handling Building once every thirty (30) hours. The personnel making rounds to the spent fuel pool are trained to be aware of the change in the status of the spent fuel pool.

They are instructed to check the temperature and level in the pool and conditions around the pool during plant rounds.

4. Administrative are placed on some of the potential dilution paths.

5. The current administrative limit on spent fuel pit boron concentration is a minimum of 2,300 ppm.

6. The proposed Technical Specifications associated with the use of soluble boron credit will require spent fuel pool boron concentration to be verified on a frequency commensurate with the results of this analysis.

3.6.4 Inputs

The dilution sources and flow rates identified and discussed in the SFP boron dilution analysis are summarized in Table 12.

Table 12 Potential Boron Dilution Sources and Approximate Flow Rates Systems Potential Boron Dilution Sources Approximate Flow Rates (GPM)

BRS Recycle Holdup Tanks to SFP Transfer Canal 100

CVCS Primary Water Storage Tank Pumps to SFP via the Boric Acid Blender 205 Primary Water Storage Tanks pumps to SFP via the connection downstream of the spent fuel pit demineralizer filter 220 Primary Water Storage Tanks pumps to SFP Primary Water System via the spent fuel pit demineralizer resin 205 flushing connection for Braidwood only, not for Byron Primary Water Storage Tanks pumps to SFP via the 2" station in the spent fuel pit area 256 Condensate storage tank and flushing pump to Demineralizer Water System SFP via one of the 3" hose stations 420 Condensate storage tank and flushing pump to SFP via one of the 2" hose stations 340 Fire Protection System To SFP via one of the 3 fire hose stations in the spent fuel pit area 191

Station Heating System Blowdown rate from Station Heating System surge tank of 6,000 gallons 23

Spent Fuel Pit Demineralizers Spent Fuel Pit Demineralizers (1/2FC01D) 180

Page 44 of 54 ATTACHMENT 1 Evaluation of Proposed Change

3.6.5 Results and Conclusion

The time to the high-level alarm and the time of dilution to 575 ppm are listed in Table 13. Based on the dilution analysis and the results listed in Table 13, it is concluded that there is plenty of time for operator to stop an unplanned or inadvertent event which would result in the dilution of the SFP boron concentration from 2000 ppm to 575 ppm based on the following reasoning:

x 568,000 gallons of water is needed to dilute the SFP from boron concentration of 2000 ppm to 575 ppm. To provide such a substantial amount of water, an operator would have to initiate the dilution flow, then abandon monitoring of the spent fuel pit level, and ignore tagged valves, administrative procedures, and a high-level alarm. 568,000 gallons of water is more than the SFP water volume corresponding to the low-level alarm setpoint elevation. It would take more than 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> to complete the dilution for a worse case scenario.

x Since such a substantial amount of water turnover is required, a spent fuel pool dilution event would be readily detected by plant personnel via alarms, flooding in the fuel handling building or by normal operator rounds through the SFP area at least once per thirty (30) hours.

x It should be noted that the SFP storage rack keff will be less than 1.0 with a 95% probability at a 95% confidence level with the SFP filled with non-borated water.

Page 45 of 54 ATTACHMENT 1 Evaluation of Proposed Change

Table 13 Times to the High-Level Alarm and Times of dilution for Potential Boron Dilution Events Systems Potential Boron Dilution Sources Time to High Level Time of Dilution to Alarm (minutes) 575 ppm (hours)

BRS Recycle Holdup Tanks to SFP Transfer Canal Not Credible Not Credible

CVCS Primary Water Storage Tank Pumps to SFP via the Boric Acid Blender 43.02 46.21 Primary Water Storage Tanks pumps to SFP via the connection 40.09 43.06 downstream of the spent fuel pit demineralizer filter Primary Water Storage Tanks pumps Primary Water to SFP via the spent fuel pit System demineralizer resin flushing 43.02 46.21 connection for Braidwood only, not for Byron Primary Water Storage Tanks pumps to SFP via the 2" station in the spent 34.45 37.01 fuel pit area Condensate storage tank and flushing pump to SFP via one of the 3" hose 21.00 22.56 Demineralizer stations Water System Condensate storage tank and flushing pump to SFP via one of the 2" hose 25.94 27.86 stations To SFP via one of the 3 fire hose Fire Protection stations in the spent fuel pit area 46.18 49.60 System A break in one of the fire protection hose station supply lines during Not Credible Not Credible Seismic Events Station Heating Blowdown from Station Heating System System surge tank of 6,000 gallons. Not Credible Not Credible Total volume in 30 hrs, 41,400 gallons Spent Fuel Pit Spent Fuel Pit Demineralizers Demineralizers 1/2FC01D Not Credible Not Credible

3.7 Other Analyses (T-H Qualification)

The "Spent Fuel Pool Temperature Analysis" calculations (BRW-00-0010-M / BYR2000-007 Rev.

000AA) were revised to re-evaluate the peak and average fuel heat fluxes for the new Framatome GAIA fuel assemblies. These heat fluxes are crucial for determining water temperature between pins and the maximum cladding temperature. This assessment also involves comparing the maximum heat flux against the Departure from Nucleate Boiling (DNB) critical heat flux in calculation HI-982085, Rev. 5/5A, which ensures that local water remains subcooled.

The conclusions drawn from BRW-00-0010-M / BYR2000-007 Rev. 000AA indicate that the new GAIA fuel assemblies exhibit lower average and peak heat fluxes per assembly across all analyzed cases. This difference can be attributed to the greater heat transfer surface area of GAIA fuel compared to the existing Westinghouse fuel. The heat flux decay heat load inputs remain

Page 46 of 54 ATTACHMENT 1 Evaluation of Proposed Change

consistent for the new fuel according to the ASB 9-2 methodology. These findings affirm that the new GAIA fuel will indeed possess lower average and peak heat fluxes. Furthermore, various parameters outlined in a previous calculation revision still serve as bounding values for the new GAIA fuel:

a) Maximum Spent Fuel Pit (SFP) temperature b) Time to boil upon loss of SFP cooling c) Maximum fuel cladding temperature d) Maximum fuel heat flux e) Vaporization rate if boiling initiates f) Maximum local water temperature and fuel cladding temperatures in UFSAR Table 9.1-5.

4.0 REGULATORY EVALUATION

4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA

The regulatory requirements associated with this amendment application include the following:

Appendix A to Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50), General Design Criterion (GDC) 61, "Fuel storage and handling and radioactivity control, " requires that fuel storage systems be designed to assure adequate safety under normal and postulated accident conditions. The criticality safety evaluation demonstrates continued conformance with GDC 61.

No administrative or physical changes are proposed that affect the ability to perform inspections and testing, shielding for radiation protection, confinement and filtering of potential effluents, or decay heat removal, nor is there any impact on assumed fuel storage coolant inventory under accident conditions.

10 CFR 50 Appendix A, GDC 62, "Prevention of criticality in fuel storage and handling," states that "criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations." The criticality safety evaluation demonstrates the continued conformance with GDC 62. The proposed storage practices are not appreciably different than those currently employed to store fuel assemblies in the SFP.

10 CFR 50.68, "Criticality accident requirements," subpart (b), regarding fresh fuel storage specifies that:

" (2) the estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used. "

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" (3) If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level.

This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used. "

10 CFR 50.68, "Criticality accident requirements," subpart (b), regarding spent fuel storage specifies that:

"(4) if credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water."

The updated criticality safety analyses performed demonstrates that the 10 CFR 50.68(b)(4) criteria are met.

10 CFR 50.36, "Technical specifications, " details the content and information that must be included in a station's Technical Specifications (TS). In accordance with 10 CFR 50.36, TSs are required to include (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 spent fuel storage racks are design features which are included in the Braidwood and Byron TS in accordance with 10 CFR 50.36. The proposed changes to the Braidwood and Byron TS would ensure that the design features relied upon in the criticality safety analysis (i.e., the storage configuration within the spent fuel storage racks) are properly described in the TSs. The proposed TS changes are consistent with the format, level of detail, and structure of NUREG-1431, Standard Technical Specifications Westinghouse Plants, Volume 1 Specifications, Revision 5.0 dated March 2021.

The following regulatory requirements pertinent to this amendment application are unaffected by the proposed changes:

x 10 CFR 50 Appendix A, GDC Criterion 1, "Quality Standards and Records" x 10 CFR 50 Appendix A, GDC Criterion 2, "Design Bases for Protection Against Natural Phenomena" x 10 CFR 50 Appendix A, GDC Criterion 3, "Fire Protection" x 10 CFR 50 Appendix A, GDC Criterion 4, "Environmental and Dynamic Effects Design Bases" x 10 CFR 50 Appendix A, GDC Criterion 5, "Sharing of Structures, Systems and Components" x 10 CFR 50 Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants"

Page 48 of 54 ATTACHMENT 1 Evaluation of Proposed Change

In conclusion, based on considerations discussed herein, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

4.2 PRECEDENTS

The NRC has approved similar submittals as indicated below:

1) NRC Safety Evaluation Report, Callaway Plant, Unit No. 1 - Issuance of Amendment No.

232 Regarding Technical Specification Changes for Spent Fuel Storage (EPID L-2022-LLA-0132), dated May 10, 2023 (ADAMS Accession No. ML23093A095).

4.3 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit, " Constellation Energy Generation, LLC (CEG) requests amendments to the Technical Specifications (TS) for Braidwood Station, Units 1 and 2 (Braidwood) and Byron Station, Units 1 and 2 (Byron).

The proposed amendment is to change:

x Technical Specifications (TS) 3.7.15, "Spent Fuel Pool Boron Concentration" to increase the required spent fuel pool boron concentration to be > 2000 ppm.

x TS 3.7.16, "Spent Fuel Assembly Storage" to update figure 3.7.16-1 to include fuel from Framatome and Westinghouse.

x TS 4.3.1.b, "Fuel Storage," "Criticality" replaced with keff < 1.00, at a 95% probability, 95%

confidence level, if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1.2.3.1 of the Updated Final Safety Analysis Report (UFSAR) and keff < 0.95, at a 95% probability, 95% confidence level, if fully flooded with water borated to 550 ppm, which includes allowances for biases and uncertainties as described in Sections 9.1.2.3.5 and 9.1.3.1 of the UFSAR.

x TS 4.3.1.c and d, "Fuel Storage," "Criticality" replace "For Holtec spent fuel pool storage racks, a" with "A". (Braidwood only)

CEG performed a criticality safety evaluation for fuel assembly storage in the Braidwood and Byron Spent Fuel Pool (SFP) storage racks and New Fuel Storage Vaults to support the proposed TS change using a methodology described in Attachment 1.

Page 49 of 54 ATTACHMENT 1 Evaluation of Proposed Change

According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated;

2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3) Involve a significant reduction in a margin of safety.

In support of this determination, an evaluation of each of the three criteria set forth in 10 CFR 50.92 is provided below:

1. Does the Proposed Change Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated?

Response: No

The proposed license amendment would revise the Technical Specifications (TS) to reflect the results of an updated analysis for the storage of spent fuel at Braidwood and Byron. The updated analysis will: 1) revise the current analysis based on the latest methodologies consistent with current NRC guidance and expectations and 2) provide an evaluation that encompasses a future fuel design. Specifically, the updated analysis adopts currently accepted computational methods, assumptions, and limitations to demonstrate compliance with regulatory standards as described in RG 1.240. The resultant changes include the adoption of an analysis that credits soluble boron, and demonstrates compliance to the acceptance criteria given in 10 CFR 50.68(b)(4).

The proposed changes will not affect plant equipment or structures, including the SFP, the Spent Fuel Racks, or fuel handling equipment, nor do they affect how the equipment is operated and maintained. There are no changes to the equipment for fuel handling or how fuel assemblies are handled, including how fuel assemblies are inserted into and removed from the Spent Fuel Rack storage locations. There is no change to the administrative means for verifying correct fuel assembly storage in the SFP, or to the required response to a fuel assembly misload or drop event. There are no changes to how rod cluster control assemblies (RCCAs) are handled, including how RCCAs are inserted into or removed from a fuel assembly or other location such as an SFP storage location. Also, since the proposed changes do not modify plant equipment or its operation and maintenance, including equipment used to maintain SFP soluble boron levels, the proposed changes will not increase the likelihood of a boron dilution event or the plant response to mitigate one should it occur. Thus, the probability of a fuel assembly misloading or a fuel assembly drop in the SFP will not significantly increase due to the proposed changes.

Several postulated accidents for the SFP were reviewed for the proposed changes, which included postulated fuel assembly misload scenarios. The criticality safety evaluation for the SFP concluded that the limiting accident, which bounds the other scenarios, is the incorrect application of the fuel loading curve resulting in multiple misloaded fuel assemblies in the Region 2 fuel

Page 50 of 54 ATTACHMENT 1 Evaluation of Proposed Change

storage region. The criticality safety evaluation concluded that an SFP soluble boron concentration of 2000 ppm will maintain keff 0.95, including uncertainties and biases, for this postulated scenario, and therefore, the TS required minimum soluble boron concentration of 2000 ppm (as proposed) will provide significant margin. Braidwood and Byron have maintained SFP soluble boron concentration greater than this value for many years, so the proposed changes will not affect the routine maintenance of the boron concentration.

As noted above, there are no changes to plant equipment, including its operation and maintenance, as a result of the proposed changes. This includes equipment associated with maintaining SFP soluble boron concentration within its limit or possible flow paths that could contribute to a boron dilution event. Thus, no new avenues for a boron dilution event will be created. There is no change regarding how the plant maintains boron concentration or responds to a boron dilution event. The criticality safety evaluation for the postulated boron dilution event shows the SFP maintains keff 0.95, including uncertainties and biases, at 550 ppm soluble boron concentration. Thus, there is no significant increase in the probability or consequences of a boron dilution accident.

The Braidwood and Byron SFPs are currently licensed to store 2984 fuel assemblies each. These maximum storage limits are unchanged.

In each of the above scenarios, the proposed changes do not significantly increase the probability of an accident previously evaluated, as the required keff margin is shown to be maintained.

Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the Proposed Change Create the Possibility of a New or Different Kind of Accident from any Accident Previously Evaluated?

Response: No

The proposed changes involve no change to any plant equipment, including how equipment is operated and maintained. There is no change to equipment used to handle fuel assemblies (or any heavy load) over the SFP, and there is no change regarding how the fuel assemblies are stored, inserted into, and removed from fuel storage locations. There is no change to how RCCAs will be inserted into or removed from a fuel assembly or other location, or otherwise how RCCAs are handled. There is no change in the manner in which SFP boron concentration is measured or maintained, as compliance with the specified TS limit would continue to be required. Thus, no new accidents are required to be postulated beyond the existing postulated accidents of a fuel misloading event or a fuel assembly drop in the SFP. There is no change in the application of the double contingency principle such that a new combination of conditions could be postulated resulting in a new or different kind of accident.

Since the proposed changes do not involve changes to plant equipment, including fuel/RCCA handling equipment or how fuel assemblies and RCCAs are handled and stored, there is no

Page 51 of 54 ATTACHMENT 1 Evaluation of Proposed Change

mechanism for creating a new or different kind of accident not previously evaluated. Therefore, it is concluded that the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the Proposed Change Involve a Significant Reduction in a Margin of Safety?

Response: No

The margin of safety is established through equipment design, operating parameters, and the setpoints at which automatic actions are initiated. The proposed changes do not alter equipment design or the way in which the equipment is operated or maintained in regard to the SFP and fuel handling activities involving the SFP. The operating parameters for the spent fuel pool are not altered by the proposed change. Further, there are no automatic actions associated with the spent fuel pool or other automatic actions within the scope of the proposed amendment. The licensing requirement for the SFP is that keff UHPDLQV 0.95 under normal and postulated accident conditions (with credit for soluble boron). The criticality safety evaluation completed in support of the proposed amendment concluded that this requirement is met. The analyses apply to all of the fuel assemblies currently stored in the Braidwood and Byron SFPs and to the future anticipated fuel design.

In addition, the criticality safety evaluation concludes that the SFP will maintain keff < 1.0 with 0 ppm soluble boron in the SFP under normal conditions, with the maximum allowed reactivity fuel assembly stored in each fuel storage location.

The criticality safety evaluation also allows the following storage configurations:

• Storing non-fuel components in any spent fuel rack storage location where fuel assemblies are allowed.

• Storing non-fuel components in the guide tubes of any fuel assembly.

For each analyzed case, the storage configuration does not increase reactivity, thus ensuring that keff margin is maintained.

Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety.

In consideration of all of the above, CEG concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and on that basis, a finding of "no significant hazards consideration" is justified.

Page 52 of 54 ATTACHMENT 1 Evaluation of Proposed Change

4.4 CONCLUSION

S

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 proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

The proposed change 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 change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change 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 change.

6.0 REFERENCES

6.1. Licensing Evaluation References 6.1.1. NRC Safety Evaluation Report, "Byron and Braidwood - Issuance of Amendments on Spent Fuel Storage Racks", dated March 1, 2000. (ADAMS Accession No. ML003692088)

6.1.2. NRC Document, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14, 1978. (ADAMS Accession No. ML19326A787)

6.1.3. 10CFR50.68 "Criticality Accident Requirements ", dated November 16, 2006.

6.1.4. NEI 12-16, Revision 4, "Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants", dated September 30, 2019. (ADAMS Accession No. ML19269E069)

6.1.5. NRC Regulatory Guide 1.240, "Fresh and Spent Fuel Pool Criticality Analyses", dated March 2021. (ADAMS Accession No. ML20352A361)

6.1.6. Certificate of Compliance, Certificate No. 9319 for MAP-12 and MAP-13, dated November 30, 2022. (ADAMS Accession No. ML23033A348)

6.1.7. ANP-10353P-A, Revision 0, "Increased Enrichment for PWRs", dated March 2023.

(ADAMS Accession No. ML23139A274)

Page 53 of 54 ATTACHMENT 1 Evaluation of Proposed Change

6.1.8. NRC Safety Evaluation Report, "Issuance of Amendments - Byron and Braidwood Stations (TAC NOS. M99170, M99171, M99168, and M99169) ", dated December 4, 1997. (ADAMS Accession No. ML020870650)

6.1.9. HI-982094, Revision 5, "Criticality Evaluation for the Byron/Braidwood Rack Installation Project", dated December 2013.

6.1.10. NRC Safety Evaluation Report, "Callaway Plant, Unit No. 1 - Issuance of Amendment No.

232 Regarding Technical Specification Changes for Spent Fuel Storage (EPID L-2022-LLA-0132) ", dated May 10, 2023 (ADAMS Accession No. ML23093A095).

6.2. Benchmarking / Criticality Analyses References

6.2.1. NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology, " dated January 2001. (ADAMS Accession No. ML010870155)

6.2.2. R. Ferrer, J. Hykes, H. Akkurt, R. Hall, "Extension of Reactivity Decrement Uncertainty to Advanced PWR Fuels via Stochastic Sampling and Sensitivity-Based Verification, " M&C 2023 - The International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, August 13-17, 2023, Niagara Falls, Ontario, Canada.

6.2.3. BRW-22-0026-N Revision 0 and BYR22-018 Revision 0, "Criticality Calculation Benchmarking with MCNP V6.2", dated August 2023.

6.2.4. BRW-22-0027-N Revision 0 and BYR22-019 Revision 0, "Braidwood-Byron Criticality Analysis for New Fuel Vault", dated August 2023.

6.2.5. BRW-22-0028-N Revision 0 and BYR22-020 Revision 0, "Braidwood-Byron Criticality Analysis for Spent Fuel Pool", dated September 2023.

6.2.6. NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses", dated March 2003. (ADAMS Accession No. ML031110292)

Page 54 of 54 ATTACHMENT 2

BRAIDWOOD STATION UNITS 1 and 2

Renewed Facility Operating License Nos. NPF-72 and NPF-77

Docket Nos. STN-50-456 and STN-50-457

Mark-up of Technical Specifications Pages

Spent Fuel Pool Boron Concentration 3.7.15

3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Boron Concentration 2000

LCO 3.7.15 The spent fuel pool boron concentration shall be ! 300 ppm.

APPLICABILITY: Whenever fuel assemblies are stored in the spent fuel pool.

ACTIONS

---

NOTE-------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME

A. Spent fuel pool boron A.1 Suspend movement of Immediately concentration not fuel assemblies in within limit. the spent fuel pool.

AND

A.2 Initiate action to Immediately restore spent fuel pool boron concentration to within limit.

SURVEILLANCE REQUIREMENTS

SURVEILLANCE FREQUENCY

SR 3.7.15.1 Verify the spent fuel pool boron concentration is within limit. In accordance with the Surveillance Frequency Control Program

BRAIDWOOD UNITS 1 & 2 3.7.15 1 Amendment 165 Spent Fuel Assembly Storage 3.7.16

Figure being replaced by figure on next page

Figure 3.7.16-1 (page 1 of 1)

Region 2 Fuel Assembly Burnup Requirements (Holtec Spent Fuel Pool Storage Racks)

BRAIDWOOD UNITS 1 & 2 3.7.16 3 Amendment 145

Design Features 4.0

4.0 DESIGN FEATURES

4.3 Fuel Storage

4.3.1 Criticality

The spent fuel storage racks are designed and shall be maintained, as applicable, with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b. For Holtec spent fuel pool storage racks, keff ! 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Holtec International Report HI-982094, "Criticality Analysis for Byron/Braidwood Rack Installation Project," Project No. 80944, 1998;

c. For Holtec spent fuel pool storage racks, a nominal 10.888 A inch north-south and 10.574 inch east-west center to center distance between fuel assemblies placed in Region 1 racks; and

d. For Holtec spent fuel pool storage racks, a nominal 8.97 A inch center to center distance between fuel assemblies placed in Region 2 racks.

4.3.2 Drainage The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 410 ft, 0 inches.

4.3.3 Capacity

The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 2984 fuel assemblies.

keff < 1.00, at a 95% probability, 95% confidence level, if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1.2.3.1 of the Updated Final Safety Analysis Report (UFSAR) and keff < 0.95, at a 95% probability, 95% confidence level, if fully flooded with water borated to 550 ppm, which includes allowances for biases and uncertainties as described in Sections 9.1.2.3.5 and 9.1.3.1 of the UFSAR.

BRAIDWOOD UNITS 1 & 2 4.0 2 Amendment 193 ATTACHMENT 3

BYRON STATION UNITS 1 and 2

Renewed Facility Operating License Nos. NPF-37 and NPF-66

Docket Nos. STN-50-454 and STN-50-455

Mark-up of Technical Specifications Page

Spent Fuel Pool Boron Concentration 3.7.15

3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Boron Concentration 2000

LCO 3.7.15 The spent fuel pool boron concentration shall be! 300 ppm.

APPLICABILITY: Whenever fuel assemblies are stored in the spent fuel pool.

ACTIONS

---

NOTE-------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME

A. Spent fuel pool boron concentration not A.1 Suspend movement of fuel assemblies in Immediately within limit. the spent fuel pool.

AND

A.2 Initiate action to Immediately restore spent fuel pool boron concentration to within limit.

BYRON UNITS 1 & 2 3.7.15 1 Amendment 165 Spent Fuel Assembly Storage 3.7.16 Figure being replaced by figure on next page

INITIAL U-235 ENRICHMENT (w/o)

Figure 3.7.16-1 (page 1 of 1)

Region 2 Fuel Assembly Burnup Requirements

BYRON UNITS 1 & 2 3.7.16 3 Amendment 165

Design Features 4.0

4.0 DESIGN FEATURES

4.2 Reactor Core (continued)

4.2.2 Control Rod Assemblies

The reactor core shall contain 53 control rod assemblies. The control material shall be silver indium cadmium, hafnium, or a mixture of both types.

4.3 Fuel Storage

4.3.1 Criticality

The spent fuel storage racks are designed and shall be maintained, as applicable, with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b. A keff 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Holtec International Report HI-982094, "Criticality Analysis for Byron/Braidwood Rack Installation Project," Project No.

80944, 1998;

c. A nominal 10.888 inch north-south and 10.574 inch east-west center to center distance between fuel assemblies placed in Region 1 racks; and

d. A nominal 8.97 inch center to center distance between fuelassemblies placed in Region 2 racks.

4.3.2 Drainage

The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 410 ft, 0 inches.

4.3.3 Capacity

The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 2984 fuel assemblies.

keff < 1.00, at a 95% probability, 95% confidence level, if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.1.2.3.1 of the Updated Final Safety Analysis Report (UFSAR) and keff < 0.95, at a 95% probability, 95% confidence level, if fully flooded with water borated to 550 ppm, which includes allowances for biases and uncertainties as described in Sections 9.1.2.3.5 and 9.1.3.1 of the UFSAR.

BYRON UNITS 1 & 2 4.0 2 Amendment 206/207 ATTACHMENT 4

Affidavit for withholding Constellation Energy Generation, LLC for Attachment 8

AFFIDAVIT

1. My name is Kevin Lueshen. I am Senior Manager, Licensing for Constellation Energy Generation, LLC (CEG) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by CEG to determine whether certain CEG information is proprietary. I am familiar with the policies established by CEG to ensure the proper application of these criteria.

3. I am familiar with the CEG information contained in Attachment 8 to CEG letter RS-23-093 dated September 29, 2023, with subject "License Amendment to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, "Spent Fuel Pool Boron Concentration", 3.7.16, "Spent Fuel Assembly Storage, 4.3.1 "Fuel Storage, Criticality" and referred to herein as "Document. " Information contained in this Document has been classified by CEG as proprietary in accordance with the policies established by CEG for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by CEG and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information. "

6. The following criteria are customarily applied by CEG to determine whether information should be classified as proprietary:

(a) The information reveals details of CEGs research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for CEG.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for CEG in product optimization or marketability.

(e) The information is vital to a competitive advantage held by CEG, would be helpful to competitors to CEG, and would likely cause substantial harm to the competitive position of CEG. The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(c), 6(d), and 6(e) above.

7. In accordance with CEGs policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside CEG only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. CEG policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

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

Executed on: (9/29/2023)

Digitally signed Lueshe by Lueshen, Kevin n, Kevin Date: 2023.09.29 10:18:56 -05'00'

Kevin.Lueshen@Constellation.com 4300 Winfield Road Warrenville, IL 60555

ATTACHMENT 5

Affidavit for withholding Framatome for Attachment 8

A F F I D A V I T

1. My name is Morris Byram. I am Manager, Licensing & Regulatory Affairs for

Framatome Inc. (Framatome) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by Framatome to determine whether

certain Framatome information is proprietary. I am familiar with the policies established by

Framatome to ensure the proper application of these criteria.

3. I am familiar with the Framatome information contained in Attachment 7

to Constellation letter RS-23-093 dated September 29, 2023, with subject License Amendment

to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Technical Specifications 3.7.15, "Spent Fuel Pool Boron Concentration, 3.7.16, "Spent Fuel Assembly Storage, 4.3.1

Fuel Storage, Criticality, and referred to herein as Document. Information contained in these

Document has been classified by Framatome as proprietary in accordance with the policies

established by Framatome for the control and protection of proprietary and confidential

information.

4. This Document contains information of a proprietary and confidential nature

and is of the type customarily held in confidence by Framatome and not made available to the

public. Based on my experience, I am aware that other companies regard information of the

kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory

Commission in confidence with the request that the information contained in this Document be

withheld from public disclosure. The request for withholding of proprietary information is made in

accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial

information.

6. The following criteria are customarily applied by Framatome to determine

whether information should be classified as proprietary:

(a) The information reveals details of Framatomes research and development

plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to

significantly reduce its expenditures, in time or resources, to design, produce,

or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a

process, methodology, or component, the application of which results in a

competitive advantage for Framatome.

(d) The information reveals certain distinguishing aspects of a process,

methodology, or component, the exclusive use of which provides a

competitive advantage for Framatome in product optimization or marketability.

(e) The information is vital to a competitive advantage held by Framatome, would

be helpful to competitors to Framatome, and would likely cause substantial

harm to the competitive position of Framatome.

The information in this Document is considered proprietary for the reasons set forth in

paragraph 6(b) and 6(d) above.

7. In accordance with Framatomes policies governing the protection and control

of information, proprietary information contained in this Document has been made available, on

a limited basis, to others outside Framatome only as required and under suitable agreement

providing for nondisclosure and limited use of the information.

8. Framatome policy requires that proprietary information be kept in a secured

file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge,

information, and belief.

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

Executed on: (9/25/2023)

(NAME) morris.byram@framatome.com

2101 Horn Rapids Road Richland, WA 99354 ATTACHMENT 6

NEI 12-16 checklists

Compliance with NEI 12-16, Benchmarking Calculations (Reference 6.2.3):

The section numbers referenced in the following table are specific to the Benchmarking Calculation Design Analysis (Reference 6.2.3).

Subject Included Notes / Explanation Code validation methodology and bases Yes Section 5 New Fuel Yes Sections 2.1.1 and 2.1.2 Depleted Fuel Yes Sections 2.1.1 and 2.1.3 MOX Yes Sections 2.1.1 and 2.1.3 HTC Yes Section 2.1.3 Convergence Yes Section 7 Trends Yes Sections 8 and 9 Bias and uncertainty Yes Sections 8 and 9 Range of applicability Yes Section 5.2 Analysis of Area of Applicability coverage Yes Section 5.2

Compliance with NEI 12-16, NFV Criticality Calculations (Reference 6.2.4):

The section, table and figure numbers referenced in the following table are specific to the NFV Criticality Calculation Design Analysis (Reference 6.2.4).

Subject Included Notes / Explanation

1. Introduction and Overview Purpose of submittal Yes Section 1 A new analysis is performed Changes requested Yes introducing a new fuel (GAIA 17x17) and a future fuel (AFM-GAIA 17x17).

Summary of physical changes No Not applicable.

Summary of Tech Spec changes No Provided in LAR.

Summary of analytical scope Yes Section 1

2. Acceptance Criteria and Regulatory Guidance Summary of requirements and guidance Yes Section 2 Requirements documents referenced Yes Section 2 Guidance documents referenced Yes Section 2 Acceptance criteria described Yes Section 2

3. Reactor and Fuel Design Description Describe reactor operating parameters No Not applicable.

Describe all fuel in pool No Not applicable.

Geometric dimensions (Nominal and Tolerances) No Not applicable.

Schematic of guide tube patterns No Not applicable.

Material compositions No Not applicable.

Describe future fuel to be covered Yes Section 3 Geometric dimensions (Nominal and Tolerances) Yes Table 3.1 Schematic of guide tube patterns Yes Figure 7.1 Material compositions Yes Table 3.3 Describe all fuel inserts No Not applicable.

Geometric Dimensions (Nominal and Tolerances) No Not applicable.

Schematic (axial/cross-section) No Not applicable.

Material compositions No Not applicable.

Describe non-standard fuel No Not applicable.

Geometric dimensions No Not applicable.

Describe non-fuel items in fuel cells No Not applicable.

Nominal and tolerance dimensions No Not applicable.

4. Spent Fuel Pool/Storage Rack Description New fuel vault & Storage rack description Yes NFV is analyzed.

Nominal and tolerance dimensions Yes Section 7.2, Table 3.2 Schematic (axial/cross-section) Yes Figures 7.3 and 7.4 Material compositions Yes Table 3.3 Spent fuel pool, Storage rack description No Not applicable.

Nominal and tolerance dimensions No Not applicable.

Schematic (axial/cross-section) No Not applicable.

Material compositions No Not applicable.

Other Reactivity Cont rol Devices (Inserts) No Not applicable.

Nominal and tolerance dimensions No Not applicable.

Schematic (axial/cross-section) No Not applicable.

Material compositions No Not applicable.

5. Overview of the Method of Analysis

New fuel rack analysis description Yes NEI 12-16, RG 1.240, using MCNP 6.2

Storage geometries Yes Section 7.2 Bounding assembly design(s) Yes Section 7.1 Integral absorber credit No Not applicable.

Accident analysis Yes Section 7 Spent fuel storage rack analysis description No Not applicable.

Storage geometries No Not applicable.

Bounding assembly design(s) No Not applicable.

Soluble boron credit No Not applicable.

Boron dilution analysis No Not applicable.

Burnup credit No Not applicable.

Decay/Cooling time credit No Not applicable.

Integral absorber credit No Not applicable.

Other credit No Not applicable.

Fixed neutron absorbers No Not applicable.

Aging management program No Not applicable.

Accident analysis No Not applicable.

Temperature increase No Not applicable.

Assembly drop No Not applicable.

Single assembly misload No Not applicable.

Multiple misload No Not applicable.

Boron dilution No Not applicable.

Other No Not applicable.

Fuel out of rack analysis No Not applicable.

Handling No Not applicable.

Movement No Not applicable.

Inspection No Not applicable.

6. Computer Codes, Cross Sections and Validation Overview Code/Modules Used for Calculation of k eff Yes MCNP 6.2 is used.

Cross section library Yes Section 6 Description of nuclides used Yes Table 3.3 Convergence checks Yes Section 6 Code/Module Used for Depletion Calculation No Not applicable.

Cross section library No Not applicable.

Description of nuclides used No Not applicable.

Convergence checks No Not applicable.

Validation of Code and Library Yes Section 5 Major Actinides and Structural Materials No Not applicable.

Minor Actinides and Fission Products No Not applicable.

Absorbers Credited No Not applicable.

7. Criticality Safety Analysis of the New Fuel Rack Rack model Yes Boundary conditions Yes Section 7.2 Source distribution Yes Section 6 Geometry restrictions Yes Section 7.2, Table 3.2 Limiting fuel design Yes Section 7.1 Fuel density Yes Section 8.2 Burnable Poisons Yes Only in sensitivity analysis Fuel dimensions Yes Section 8.4 Axial blankets No Not modeled.

Limiting rack model Section 8.5 Storage vault dimensions and materials Yes Attachment C Temperature Yes Attachment C Multiple regions/configurations No Not applicable.

Flooded Yes Section 8.1 Low density moderator Yes Section 8.1 Eccentric fuel placement Yes Section 8.3 Tolerances Yes Fuel geometry Yes Table 8.3 Fuel pin pitch Yes Table 8.3 Fuel pellet OD Yes Table 8.3 Fuel clad OD Yes Table 8.3 Fuel content Yes Table 8.3 Enrichment Yes Table 8.3 Density Yes Table 8.3 No Conservatively modeled in a Integral absorber sensitivity analysis described in Attachment C.

Rack geometry Yes Table 8.4 Rack pitch Yes Table 8.4 Cell wall thickness Yes Table 8.4 Storage vault dimensions/materials Yes Table 8.4, Attachment C Code uncertainty Yes Table 3.4 Biases Yes Table 3.5 Temperature Yes Table 8.5 Code bias Yes Table 8.5 Moderator Conditions Yes Section 8.1 Fully flooded and optimum density moderator Yes Section 8.1

8. Depletion Analysis for Spent Fuel No Not applicable Depletion Model Considerations No Not applicable Time step verification No Not applicable Convergence verification No Not applicable Simplifications No Not applicable Non-uniform enrichments No Not applicable Post Depletion Nuclide Adjustment No Not applicable Cooling Time No Not applicable Depletion Parameters No Not applicable Burnable Absorbers No Not applicable Integral Absorbers No Not applicable Soluble Boron No Not applicable Fuel and Moderator Temperature No Not applicable Power No Not applicable Control rod insertion No Not applicable Atypical Cycle Operating History No Not applicable

9. Criticality Safety Analysis of Spent Fuel applicable Pool Storage Racks No Not Rack model No Not applicable Boundary conditions No Not applicable Source distribution No Not applicable Geometry restrictions No Not applicable Design Basis Fuel Description No Not applicable Fuel density No Not applicable Burnable Poisons No Not applicable Fuel assembly inserts No Not applicable Fuel dimensions No Not applicable Axial blankets No Not applicable Configurations considered No Not applicable Borated No Not applicable Unborated No Not applicable Multiple rack designs No Not applicable Alternate storage geometry No Not applicable Reactivity Control Devices No Not applicable Fuel Assembly Inserts No Not applicable Storage Cell Inserts No Not applicable Storage Cell Blocking Devices No Not applicable Axial burnup shapes No Not applicable Uniform/Distributed No Not applicable Nodalization No Not applicable Blankets modeled No Not applicable Tolerances/Uncertainties No Not applicable Fuel geometry No Not applicable Fuel rod pin pitch No Not applicable Fuel pellet OD No Not applicable Cladding OD No Not applicable Axial fuel position No Not applicable Fuel content No Not applicable Enrichment No Not applicable Density No Not applicable Assembly insert dimensions and materials No Not applicable Rack geometry No Not applicable Flux-trap size (width) No Not applicable Rack cell pitch No Not applicable Rack wall thickness No Not applicable Neutron Absorber Dimensions No Not applicable Rack insert dimensions and materials No Not applicable Code validation uncertainty No Not applicable Criticality case uncertainty No Not applicable Depletion Uncertainty No Not applicable Burnup Uncertainty No Not applicable Biases No Not applicable Design Basis Fuel design No Not applicable Code bias No Not applicable Temperature No Not applicable Eccentric fuel placement No Not applicable Incore thimble depletion effect No Not applicable NRC administrative margin No Not applicable Modeling simplifications No Not applicable Identified and described No Not applicable

10. Interface Analysis No Not applicable Interface configurations analyzed No Not applicable Between dissimilar racks No Not applicable Between storage configurations within a rack No Not applicable Interface restrictions No Not applicable

11. Normal Conditions No During normal condition, neutrons will not be moderated.

Fuel handling equipment No Not applicable Administrative controls No Not applicable Fuel inspection equipment or processes No Not applicable Fuel reconstitution No Not applicable Not applicable

12. Accident Analysis Yes Not applicable Boron dilution No Not applicable Normal conditions No Not applicable Accident conditions No Not applicable Single assembly misload No Not applicable Fuel assembly misplacement No Not applicable Neutron Absorber Insert Misload No Not applicable Multiple fuel misload No Not applicable Dropped assembly No Not applicable Temperature Yes Attachment C Seismic event/other natural phenomena No Not applicable

13. Analysis Results and Conclusions Summary of results Yes Tables 1.1 and 8.5 Burnup curve(s) No Not applicable Intermediate Decay time treatment No Not applicable New administrative controls No Not applicable Technical Specification markups No Not applicable

14. References Yes Section 10

Compliance with NEI 12-16, SFP Criticality Calculations (Reference 6.2.5):

The section, table and figure numbers referenced in the following table are specific to the SFP Criticality Calculation Design Analysis (Reference 6.2.5).

Subject Included Notes / Explanation

1. Introduction and Overview Purpose of submittal Yes Section 1 A new analysis is performed Changes requested Yes introducing a new fuel (GAIA 17x17) and a future fuel (AFM-GAIA 17x17).

Summary of physical changes No Not applicable.

Summary of Tech Spec changes No Provided in LAR.

Summary of analytical scope Yes Section 1

2. Acceptance Criteria and Regulatory Guidance Summary of requirements and guidance Yes Section 2 Requirements documents referenced Yes Section 2 Guidance documents referenced Yes Section 2 Acceptance criteria described Yes Section 2

3. Reactor and Fuel Design Description Describe reactor operating parameters Yes Section 3.3 Describe all fuel in pool No Not applicable.

Geometric dimensions (Nominal and Tolerances) No Not applicable.

Schematic of guide tube patterns No Not applicable.

Material compositions No Not applicable.

Describe future fuel to be covered Yes Geometric dimensions (Nominal and Tolerances) Yes Table 3.1.1 Schematic of guide tube patterns Yes Figure 7.1.2 Material compositions Yes Table 3.2.3 Describe all fuel inserts No Not considered.

Geometric Dimensions (Nominal and Tolerances) No Not credited Schematic (axial/cross-section) No Not credited Material compositions No Not credited Describe non-standard fuel No Not considered.

Geometric dimensions No Not considered.

Describe non-fuel items in fuel cells No Not considered.

Nominal and tolerance dimensions No Not considered.

4. Spent Fuel Pool/Storage Rack Description New fuel vault & Storage rack description No Not applicable Nominal and tolerance dimensions No Not applicable Schematic (axial/cross-section) No Not applicable Material compositions No Not applicable Spent fuel pool, Storage rack description Yes Section 3 Nominal and tolerance dimensions Yes Table 3.2.1 Figures 7.1.1 and 7.2.1 show Schematic (axial/cross-section) Yes radial cross section. No axial variation in modeling.

Material compositions Yes Table 3.2.3 Other Reactivity Cont rol Devices (Inserts) No Not credited Nominal and tolerance dimensions No Not credited Schematic (axial/cross-section) No Not credited Material compositions No Not credited

5. Overview of the Method of Analysis New fuel rack analysis description No Storage geometries No Not applicable Bounding assembly design(s) No Not applicable Integral absorber credit No Not applicable Accident analysis No Not applicable Spent fuel storage rack analysis description Yes Section 7 Storage geometries Yes Section 7 Bounding assembly design(s) Yes Section 7 Soluble boron credit Yes

Boron dilution analysis No Boron dilution analyzed in another report.

Burnup credit Yes Section 7.2 Decay/Cooling time credit Yes Section 7.2 Integral absorber credit No Not credited Other credit No Not credited Fixed neutron absorbers Yes Section 3 Aging management program Yes Section 3 Accident analysis Yes Section 7 Temperature increase Yes Section 7 Assembly drop Yes Section 7 Single assembly misload Yes Section 7 Multiple misload Yes Section 7

Boron dilution No Boron dilution analyzed in another report.

Other No Not credited Fuel out of rack analysis Yes Section 7.2 Handling Yes Section 7.2 Movement Yes Section 7.2 Inspection Yes Section 7.2

6. Computer Codes, Cross Sections and Validation Overview Code/Modules Used for Calculation of k eff Yes Cross section library Yes Section 6 Description of nuclides used Yes Table 3.3 Convergence checks Yes Section 6 Code/Module Used for Depletion Calculation Yes Cross section library Yes Section 6.3 Description of nuclides used Yes Section 7.2 Convergence checks No Not Applicable Validation of Code and Library Yes Major Actinides and Structural Materials Yes Benchmarking report Minor Actinides and Fission Products Yes Section 7.2

Absorbers Credited Yes Section 7.2, Benchmarking report

7. Criticality Safety Analysis of the New Fuel Rack Rack model No Not applicable Boundary conditions No Not applicable Source distribution No Not applicable Geometry restrictions No Not applicable Limiting fuel design No Not applicable Fuel density No Not applicable Burnable Poisons No Not applicable Fuel dimensions No Not applicable Axial blankets No Not applicable Limiting rack model No Not applicable Storage vault dimensions and materials No Not applicable Temperature No Not applicable Multiple regions/configurations No Not applicable Flooded No Not applicable Low density moderator No Not applicable Eccentric fuel placement No Not applicable Tolerances No Not applicable Fuel geometry No Not applicable Fuel pin pitch No Not applicable Fuel pellet OD No Not applicable Fuel clad OD No Not applicable Fuel content No Not applicable Enrichment No Not applicable Density No Not applicable Integral absorber No Not applicable Rack geometry No Not applicable Rack pitch No Not applicable Cell wall thickness No Not applicable Storage vault dimensions/materials No Not applicable Code uncertainty No Not applicable Biases No Not applicable Temperature No Not applicable Code bias No Not applicable Moderator Conditions No Not applicable Fully flooded and optimum density moderator No Not applicable

8. Depletion Analysis for Spent Fuel Depletion Model Considerations Yes Time step verification Yes Section 7.2 Convergence verification No Not Applicable Simplifications Yes Section 7.2 Non-uniform enrichments Yes Section 7.2 Post Depletion Nuclide Adjustment Yes Section 7.2 Cooling Time Yes Section 7.2 Depletion Parameters Yes Burnable Absorbers Yes Section 7.2 Integral Absorbers Yes Section 7.2 Soluble Boron Yes Sections 3.3 and 7.2 Fuel and Moderator Temperature Yes Sections 3.3 and 7.2 Power Yes Sections 3.3 and 7.2 Control rod insertion Yes Section 7.2 Atypical Cycle Operating History Yes Section 7.2

9. Criticality Safety Analysis of Spent Fuel Pool Storage Racks Rack model Yes Section 7 Boundary conditions Yes Section 7 Source distribution Yes Section 6 Geometry restrictions Design Basis Fuel Description Yes Sections 3, 4 and 7 Fuel density Yes Section 3 Burnable Poisons Yes Section 3.1 Fuel assembly inserts No Not credited.

Fuel dimensions Yes Section 3 Axial blankets No Not credited Configurations considered Yes Section 7 Borated Yes Section 7 Unborated Yes Section 7 Multiple rack designs Yes Section 7 Alternate storage geometry No Not credited.

Reactivity Control Devices No Not credited.

Fuel Assembly Inserts No Not credited.

Storage Cell Inserts No Not credited.

Storage Cell Blocking Devices No Not credited.

Axial burnup shapes No Not credited.

Uniform/Distributed No Not credited.

Nodalization No Not credited.

Blankets modeled No Not credited.

Tolerances/Uncertainties Yes Fuel geometry Yes Sections 3 and 7 Fuel rod pin pitch Yes Sections 3 and 7 Fuel pellet OD Yes Sections 3 and 7 Cladding OD Yes Sections 3 and 7 Axial fuel position Yes Sections 3 and 7 Fuel content Yes Sections 3 and 7 Enrichment Yes Sections 3 and 7 Density Yes Sections 3 and 7 Assembly insert dimensions and materials No Not credited Rack geometry Yes Sections 3 and 7 Flux-trap size (width) Yes Sections 3 and 7 Rack cell pitch Yes Sections 3 and 7 Rack wall thickness Yes Sections 3 and 7 Neutron Absorber Dimensions Yes Sections 3 and 7 Rack insert dimensions and materials No Not applicable Code validation uncertainty Yes Sections 3 and 7 Criticality case uncertainty Yes Section 7 Depletion Uncertainty Yes Section 7.2 Burnup Uncertainty Yes Section 7.2 Biases Yes Design Basis Fuel design Yes Section 7 Code bias Yes Section 7 Temperature Yes Section 7 Eccentric fuel placement Yes Section 7 Incore thimble depletion effect No Not applicable NRC administrative margin No Not applicable Modeling simplifications Yes Sections 4 and 7 Identified and described Yes Sections 4 and 7

10. Interface Analysis Interface configurations analyzed Yes Section 7.3 Between dissimilar racks Yes Section 7.3 Between storage configurations within a rack Yes Section 7.3 Interface restrictions No Not applicable

11. Normal Conditions Yes Section 7.2 Fuel handling equipment Yes Section 7.2 Administrative controls Yes Section 7.2 Fuel inspection equipment or processes Yes Section 7.2 Fuel reconstitution Yes Section 7.2

12. Accident Analysis Yes Section 7.2

Boron dilution No Boron dilution analyzed in another report.

Normal conditions No Boron dilution analyzed in another report.

Accident conditions No Boron dilution analyzed in another report.

Single assembly misload Yes Section 7.2 Fuel assembly misplacement Yes Section 7.2 Neutron Absorber Insert Misload No Not applicable Multiple fuel misload Yes Section 7.2 Dropped assembly Yes Section 7.2 Temperature Yes Section 7.2 Seismic event/other natural phenomena Yes Section 7.2

13. Analysis Results and Conclusions Summary of results Yes Section 7.2 Burnup curve(s) Yes Section 7.2 Intermediate Decay time treatment Yes Section 7.2 New administrative controls No Not applicable Technical Specification markups No Not applicable

14. References Yes Section 10

ATTACHMENT 7

Summary of Regulatory Commitments

The following table identifies commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described to the NRC for the NRCs information and are not regulatory commitments.)

COMMITMENT TYPE COMMITMENT COMMITTED DATE OR ONE-PROGRAM "OUTAGE" TIME -MATIC ACTION (Y/N)

(Y/N)

• CEG (Braidwood) will relocate the To be implemented prior following fuel assemblies into the to implementation of the Y N Region 1 racks. Assembly IDs: D82U, license amendment D73U, D81U, and D77U