RS-24-098, Units 1 and 2 - Supplement to License Amendment Request to Revise Technical Specifications 3.7.15, Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel Assembly Storage, 4.3.1, Fuel Storage, Criticality

Units 1 and 2 - Supplement to License Amendment Request to Revise Technical Specifications 3.7.15, Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel Assembly Storage, 4.3.1, Fuel Storage, Criticality

ML24297A658
Person / Time
Site:	Braidwood, Byron
Issue date:	10/23/2024
From:	Steinman R Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML24297A657	List: RS-24-098, Units 1 and 2 - Supplement to License Amendment Request to Revise Technical Specifications 3.7.15, Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel Assembly Storage, 4.3.1, Fuel Storage, Criticality
References
RS-24-098
Download: ML24297A658 (1)

Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office WITHHOLD FROM DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390 Withhold Attachments 11 and 12 from public disclosure under 10 CFR 2.390.

When separated from Attachments 11 and 12, this document is decontrolled.

October 23, 2024 10 CFR 50.90 RS-24-098 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 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:

Supplement to License Amendment Request to Revise Technical Specifications 3.7.15, "Spent Fuel Pool Boron Concentration", 3.7.16, "Spent Fuel Assembly Storage", 4.3.1 "Fuel Storage, Criticality"

References:

1. Letter from K. Lueshen (Constellation Energy Generation, LLC) to U.S. NRC, "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", dated September 29, 2023 (ADAMS Accession No. ML23272A201)

2. Letter from J.S. Wiebe (U.S. NRC) to D.P. Rhoades (Constellation Energy Generation, LLC), Braidwood Station, Units 1 and 2 and Byron Station, Unit Nos. 1 and 2 - Request for Additional Information (EPID 2023-LLA-0136),

dated September 23, 2024 (ADAMS Accession No. ML24263A127)

In Reference 1 Constellation Energy Generation, LLC (CEG) requested an amendment to Renewed Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2 (Braidwood) and Renewed Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station, Units 1 and 2 (Byron). The requested change revises Technical Specifications (TSs) 3.7.15, "Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel Assembly Storage, and TS 4.3.1 "Fuel Storage, Criticality."

Reference 1 stated that Braidwood and Byron intended the criticality safety evaluation to support a fuel transition to Framatome GAIA fuel assemblies and potential future transition to 24-month cycles, Framatome GAIA AFM fuel assemblies, and increasing the fuel assembly Constellation

Withhold Attachments 11 and 12 from public disclosure under 10 CFR 2.390. When separated from Attachments 11 and 12, this document is decontrolled.

U.S. Nuclear Regulatory Commission October 23, 2024 Page 2 maximum uranium-235 (U-235) enrichment from 5.0 wt% to 6.5 wt%. However, the license amendment request (LAR) did not request a change to Braidwood and Byrons TS 4.3.1.a to raise the maximum licensed U-235 enrichment nor did the LAR request an exemption to 10 CFR 50.68(b)(7) which limits the maximum U-235 enrichment to 5.0 wt%. The basis for this original request was that the initial reloads of the GAIA fuel type will be of the baseline 18-month cycle variety (i.e., 5.0 wt%) and that a future LAR would revise the TS for the higher enrichment without needing to revise the bounding criticality calculation.

CEG is revising the request to only support the fuel transition to Framatome GAIA fuel and removing the previously requested increase in uranium-235 enrichment (i.e., the current enrichment limit of 5 wt% will be retained). Attachment 1 provides the evaluation of this change.

In Reference 2, the NRC requested additional information that is needed to complete review of the proposed change. The response to these requests for information are provided in (redacted public version) and Attachment 12 (proprietary version). Additionally, this supplement revises the previously submitted TS mark-ups (Attachment 3 and 4) as described below. Previously submitted TS mark-ups not explicitly listed below are unchanged by this supplement.

x TS 3.7.16, "Spent Fuel Assembly Storage" Figure 3.7.16-1 was updated to include fuel from Framatome and Westinghouse. This figure is revised so that the upper enrichment limit shown in the figure is 5.0 weight percent (wt%).

x TS 3.7.16, "Spent Fuel Assembly Storage" LCO a. and b. is revised to remove "Holtec" from the first line. And SR 3.7.16.2 is revised to remove "decay time". (Braidwood only)

Attachments 11 and 12 contain information that is confidential and proprietary to Holtec International (Holtec), Westinghouse Electric Company, LLC (WEC) and CEG. Accordingly, CEG requests that Attachments 11 and 12 be withheld from production in response to a Freedom of Information Act Request in accordance with 10 CFR 2.390, Paragraph (a)(4)

(confidential, proprietary, trade secrets). Attachments 2 and 10 are provided with the proprietary content in Attachment 12 and 11, respectively, redacted. Affidavits supporting the request for withholding for each organization are provided in Attachments 6, 7, and 8, respectively.

CEG has reviewed the information supporting the finding of no significant hazards consideration, and the environmental consideration that were previously provided to the NRC.

The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. In addition, the information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed revision to the amendment request.

Withhold Attachments 11 and 12 from public disclosure under 10 CFR 2.390. When separated from Attachments 11 and 12, this document is decontrolled.

U.S. Nuclear Regulatory Commission October 23, 2024 Page 3 CEG is notifying the State of Illinois of this supplement to a previous application for a change to the operating license by sending a copy of this letter and its attachments to the designated State Official in accordance with 10 CFR 50.91, "Notice for public comment; State consultation",

paragraph (b).

The regulatory commitments contained in this letter are summarized in Attachment 9. Should you have any questions concerning this letter, please contact Ms. Lisa Zurawski at 779-231-6196.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 23rd day of October 2024.

Respectfully, Rebecca L. Steinman Sr. Manager Licensing Constellation Energy Generation, LLC Attachments:

1.

Evaluation of Proposed Change (Non-Proprietary) 2.

Response to Request for Additional Information (Non-Proprietary) 3.

Braidwood Mark-up of Technical Specifications Page 4.

Byron Mark-up of Technical Specifications Page 5.

Revised NEI 12-16 Checklists 6.

Constellation Affidavit for Withholding for Attachments 11 and 12 7.

Holtec Affidavit for Withholding for Attachment 11 8.

Westinghouse Affidavit for Withholding for Attachment 11 9.

Summary of Regulatory Commitments 10.

HI-982094, Revision 6, "Criticality Evaluation for the Byron/Braidwood Rack Installation Project," dated October 2024 (Non-Proprietary) 11.

HI-982094, Revision 6, "Criticality Evaluation for the Byron/Braidwood Rack Installation Project," dated October 2024 (Proprietary) 12.

Response to Request for Additional Information (Proprietary) cc:

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

Steinman, Rebecca Lee Digitally signed by Steinman, Rebecca Lee Date: 2024.10.23 18:10:27

-05'00'

ATTACHMENT 1 Evaluation of the Proposed Change Page 1 of 15

Subject:

Supplement to 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 REFERENCES

ATTACHMENT 1 Evaluation of the Proposed Change Page 2 of 15 1.0

SUMMARY

DESCRIPTION In Ref. 4.1.1 Constellation Energy Generation, LLC (CEG) requested an amendment to Renewed Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2 (Braidwood) and Renewed Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station, Units 1 and 2 (Byron). The requested change revises Technical Specifications (TSs) 3.7.15, "Spent Fuel Pool Boron Concentration, 3.7.16, Spent Fuel Assembly Storage, and TS 4.3.1 "Fuel Storage, Criticality."

Ref. 4.1.1 stated that Braidwood and Byron intended the criticality safety evaluation to support a fuel transition to Framatome GAIA fuel assemblies and potential future transition to 24-month cycles, Framatome AFM-GAIA fuel assemblies, and increasing the fuel assembly maximum uranium-235 (U-235) enrichment from 5.0 wt% to 6.5 wt%. However, the license amendment request (LAR) did not request a change to Braidwood and Byrons TS 4.3.1.a to raise the maximum licensed U-235 enrichment nor did the LAR request an exemption to 10 CFR 50.68(b)(7) which limits the maximum U-235 enrichment to 5.0 wt%. The basis for this original request was that the initial reloads of the GAIA fuel type will be of the baseline 18-month cycle variety (i.e., 5.0 wt%) and that a future LAR would revise the TS for the higher enrichment without needing to revise the bounding criticality calculation.

CEG is revising the request to only support the fuel transition to Framatome GAIA fuel and removing the previously requested increase in uranium-235 enrichment (i.e., the current enrichment limit of 5 wt% will be retained).

Subsequently, the NRC requested additional information that is needed to complete review of the proposed change. The response to these requests for information are provided in RS-24-098 Attachment 2 (redacted public version) and Attachment 12 (proprietary version).

Additionally, this supplement revises the previously submitted TS mark-ups as described below.

Previously submitted TS mark-ups not explicitly listed below are unchanged by this supplement.

x TS 3.7.16, "Spent Fuel Assembly Storage" Figure 3.7.16-1 was updated to include fuel from Framatome and Westinghouse. This figure is revised so that the upper enrichment limit shown in the figure is 5.0 weight percent (wt%).

x TS 3.7.16, "Spent Fuel Assembly Storage" LCO a. and b. is revised to remove "Holtec" from the first line. And SR 3.7.16.2 is revised to remove "decay time". (Braidwood only) 2.0 DETAILED DESCRIPTION 2.1 Proposed Changes The proposed TS changes are provided in RS-24-098 Attachments 3 and 4 and include:

x TS 3.7.16, "Spent Fuel Assembly Storage" Figure 3.7.16-1 was updated to include fuel from Framatome and Westinghouse. This figure is revised so that the upper enrichment limit shown in the figure is 5.0 weight percent (wt%).

ATTACHMENT 1 Evaluation of the Proposed Change Page 3 of 15 x

TS 3.7.16, "Spent Fuel Assembly Storage" LCO a. and b. is revised to remove "Holtec" from the first line. And SR 3.7.16.2 is revised to remove "decay time". (Braidwood only)

2.2 Background

The current analyzed fuel types in use at Byron and Braidwood are Westinghouse Vantage 5, Vantage 5+, and Optimized Fuel Assembly (OFA). Beginning with Braidwood Unit 1 Cycle 26, 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 (limited to 5 wt% uranium-235).

CEG will submit a separate license amendment request in the future to transition to GAIA AFM fuel, 24-month cycles, or change methods.

3.0 TECHNICAL EVALUATION

3.1 New Fuel Vault Criticality Analysis The analysis, BRW-24-0031-N/BYR24-011, "Braidwood-Byron GAIA Criticality Analysis for New Fuel Vault" (Ref. 4.2.1), remains consistent with the analysis described in Section 3.3 of the original LAR submittal (Ref. 4.1.1), except that GAIA 17x17 fuel with 5.0 wt% enrichment was assumed as the base case fuel, with the corresponding updates described in this section.

Additional Monte Carlo N-Particle Transport (MCNP) code calculations were performed to determine the applicable biases and uncertainties for the full water density and the case with the bounding optimum-moderator density (which remains 0.04 g/cm3) for the 5.0 wt% base case fuel. Figure 1 shows how the keff was changed as a function of the water density. Figure 1 supersedes the information previously provided in Ref. 4.2.1, Attachments 1 and 8 Figure 4.

ATTACHMENT 1 Evaluation of the Proposed Change Page 4 of 15 Figure 1 keff as a Function of Water Density The geometry and material modeling bias (how NFV cavity was modeled compared to the actual, and the effect of NFV perimeter materials (such as concrete on reactivity)) were revised.

The discussion of concrete composition and the associated sensitivity studies is provided in the CEG response to Question 15 of Attachment 2. All other applicable biases were incorporated as described in Ref. 4.2.1.

Using the same uncertainties as discussed previously in Ref. 4.1.1 Section 3.3 and an enrichment up to 5.0 wt%, resulted in the revised kmax(95/95) data summarized in Table 1, which demonstrates that the revised values meet the applicable regulatory limit.

.J 0.1 0.01 Mode... tor Density (g/cm1)

ATTACHMENT 1 Evaluation of the Proposed Change Page 5 of 15 Table 1 kmax(95/95) Calculation Full-moderator density Base Case keff 0.89916 Biases Criticality code validation bias 0.00260 Eccentric positioning bias 0.00408 Geometry and material modeling bias 0.00011 Temperature reactivity effect bias 0.00000 Total bias 0.00679 Uncertainties Fuel manufacturing tolerances (95/95) 0.00558 Rack manufacturing tolerances (95/95) 0.00339 Criticality code validation uncertainty 0.00450 Monte Carlo calculational uncertainty (2 sigma) 0.00036 Statistical combination of uncertainties 0.00794 kmax(95/95) 0.9139 Regulatory Limit 0.95 Optimum-moderation water density Base Case keff 0.72691 Biases Criticality code validation bias 0.00260 Eccentric positioning bias 0.00083 Geometry and material modeling bias 0.03888 Temperature reactivity effect bias 0.02968 Total bias 0.07199 Uncertainties Fuel manufacturing tolerances (95/95) 0.00287 Rack manufacturing tolerances (95/95) 0.00898 Criticality code validation uncertainty 0.00450 Monte Carlo calculational uncertainty (2 sigma) 0.00030 Statistical combination of uncertainties 0.01045 kmax(95/95) 0.8094 Regulatory Limit 0.98 All calculations stated above were performed at a bounding enrichment of 5.0 wt%, with no credit for rods with gadolinium oxide.

ATTACHMENT 1 Evaluation of the Proposed Change Page 6 of 15 3.2 Region 1 Criticality Analysis The analysis, BRW-24-0032-N/BYR24-012, "Braidwood-Byron GAIA Criticality Analysis for Spent Fuel Pool" (Ref. 4.2.2), remains consistent with that described in Section 3.4 of the original LAR submittal (Ref. 4.1.1), except that GAIA 17x17 fuel with 5.0 wt% enrichment was assumed as the base case fuel, with the corresponding updates described in this section.

The same applicable biases provided in NEI 12-16 (Ref. 4.1.2) as discussed previously in Ref. 4.1.1 Section 3.4 were incorporated, except for those listed below.

x Base case fuel assembly (reactivity effect of M5 cladding material and Boral with the minimum density was evaluated).

x Poison (Boral) B-10 degradation bias.

A conservative Boron-10 loss (modelled by reducing the Boral density) was considered as the Boral B-10 degradation bias in the revised analysis (Ref. 4.2.2).

The same uncertainties were evaluated as discussed previously in Ref. 4.1.1 Section 3.4, except for Poison (Boral) coupon B-10 degradation, which is accounted for as a bias as described above.

Using Equation 1 from Ref. 4.1.1, the kmax(95/95) values for borated and unborated water were recalculated and summarized in Table 2. It is demonstrated that the kmax(95/95) values for both borated and unborated water meet the applicable regulatory limits.

ATTACHMENT 1 Evaluation of the Proposed Change Page 7 of 15 Table 2 kmax(95/95) Values for Borated and Unborated Water Borated water Base case keff 0.87206 Biases Criticality code validation bias 0.00260 Base case bias 0.00000 Moderator temperature bias 0.00000 Eccentric positioning bias 0.00000 Boral B-10 degradation bias 0.00130 Total bias 0.00390 Uncertainties Fuel manufacturing tolerances 0.00482 Rack manufacturing tolerances 0.00725 Criticality code validation uncertainty 0.00450 Monte Carlo calculational uncertainty (2 sigma) 0.00036 Statistical combination of uncertainties 0.00980 kmax(95/95) 0.8858 Regulatory Limit 0.95 Unborated water Base case keff 0.92318 Biases Criticality code validation bias 0.00260 Base case bias 0.00000 Moderator temperature bias 0.00000 Eccentric positioning bias 0.00000 Boral B-10 degradation bias 0.00110 Total bias 0.00370 Uncertainties Fuel manufacturing tolerances 0.00466 Rack manufacturing tolerances 0.00741 Criticality code validation uncertainty 0.00450 Monte Carlo calculational uncertainty (2 sigma) 0.00036 Statistical combination of uncertainties 0.00985 kmax(95/95) 0.9368 Regulatory Limit 1

There is no change to the evaluated abnormal and accident conditions described in Ref. 4.1.1 given the limitation in enrichment to 5.0 wt%, including temperature beyond the normal operating range, mislocated fuel assembly, assembly misload, and dropped fuel assembly. It is

ATTACHMENT 1 Evaluation of the Proposed Change Page 8 of 15 shown that these abnormal and accident conditions are either not credible or do not result in an increased reactivity.

3.3 Region 2 Criticality Analysis The analysis, BRW-24-0032-N/BYR24-012, "Braidwood-Byron GAIA Criticality Analysis for Spent Fuel Pool" (Ref. 4.2.2), remains consistent with that described in Ref. 4.1.1 Section 3.5, except that GAIA 17x17 fuel with 5.0 wt% enrichment was assumed as the base case fuel, with the corresponding updates described in this section.

For the determination of the maximum keff (kmax(95/95)), the same applicable biases provided in NEI 12-16 (Ref. 4.1.2) were incorporated as the original submittal, except for those listed below.

x Poison (Boral) coupon B-10 degradation uncertainty. The reactivity difference between the poison degraded case and the base case was considered as the poison B-10 degradation bias.

x Control rods insertion bias. The reactivity difference between the cases with the control rods insertion and the base case was considered as the bias.

The same uncertainties were evaluated as discussed previously in Section 3.5 of the original submittal (Ref. 4.1.1), except for those listed below.

x Poison (Boral) coupon B-10 degradation uncertainty (95/95 uncertainty of the bias) x Control rods insertion bias uncertainty (95/95 uncertainty of the bias)

The kmax(95/95) values for enrichments up to 5.0 wt% were recalculated using the same approach and Equation 1 from Ref. 4.1.1. Table 3 presents revised base case calculations used to generate the loading curve and the results of the polynomial burnup values and supersedes the information previously provided in Ref. 4.2.1, Attachments 1 and 8 Table 4. The resulting equation for the curve fits is summarized in Table 4, and graphically shown in Figure 2. Table 4 supersedes the information previously provided in Ref. 4.2.1, Attachments 1 and 8 Table 8.

Figure 2 supersedes the information previously provided in Ref. 4.2.1, Attachments 1 and 8 Figure 21.

The loading curve was validated by confirmatory calculations with all enrichments shown in Table 3 for the conditions with unborated water, and the case with 500 ppm borated water. It was confirmed the largest kmax(95/95) value is 0.98432 with fresh water and 0.92975 with 500 ppm borated water, both below the applicable regulatory limit. Consistent with NEI 12-16 (Ref. 4.1.2), 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. Also considering additional 25 ppm from soluble boron measurement uncertainty, 575 ppm soluble boron was required under normal conditions for the SFP Region 2.

ATTACHMENT 1 Evaluation of the Proposed Change Page 9 of 15 0BTable 3 kmax(95/95) Calculation for Region 2 Enrichment (wt. %)

2 2.5 3

3.5 4

4.5 5

Lower Bound Burnup (GWd/MTU) 0 5

15 20 25 30 35 Upper Bound Burnup (GWd/MTU) 5 10 20 25 30 35 40 keff at Lower Bound Burnup 0.96558 0.98109 0.98023 0.96825 0.97593 0.98118 0.98023 keff at Upper Bound Burnup 0.92306 0.94156 0.93100 0.94292 0.95222 0.95401 0.94643 Biases Criticality Code Validation Bias 0.00260 0.00260 0.00260 0.00260 0.00260 0.00260 0.00260 Fuel Depletion Related Geometry Change Bias 0.00676 0.00676 0.00676 0.00676 0.00676 0.00676 0.00676 Fission Gas Release Bias 0.00024 0.00075 0.00110 0.00126 0.00105 0.00030 0.00175 Axial Burnup Profile Bias 0.00224 0.00310 0.00000 0.00000 0.00000 0.00000 0.00000 Boral Degradation Bias 0.00263 0.00263 0.00263 0.00263 0.00263 0.00263 0.00263 Control Rods Insertion Bias 0.00312 0.00312 0.00312 0.00312 0.00312 0.00312 0.00312 Total Bias 0.01759 0.01896 0.01621 0.01637 0.01616 0.01541 0.01686 Uncertainties MCNP calculational uncertainty (2 sigma) 0.00064 0.00066 0.00066 0.00068 0.00072 0.00068 0.00072 Fuel Manufacturing Tolerances Uncertainty 0.00963 0.00963 0.00963 0.00963 0.00963 0.00963 0.00963 Rack Manufacturing Tolerances Uncertainty 0.00492 0.00492 0.00492 0.00492 0.00492 0.00492 0.00492 Criticality Code Validation Uncertainty 0.00450 0.00450 0.00450 0.00450 0.00450 0.00450 0.00450 Fuel Depletion Related Geometry Change Uncertainty 0.00138 0.00138 0.00138 0.00138 0.00138 0.00138 0.00138 Fission Gas Release Uncertainty 0.00089 0.00095 0.00098 0.00098 0.00103 0.00096 0.00102 Boral Degradation Uncertainty 0.00100 0.00100 0.00100 0.00100 0.00100 0.00100 0.00100

ATTACHMENT 1 Evaluation of the Proposed Change Page 10 of 15 0BTable 3 kmax(95/95) Calculation for Region 2 Control Rods Insertion Bias Uncertainty 0.00098 0.00098 0.00098 0.00098 0.00098 0.00098 0.00098 Failed Cladding Uncertainty 0.00324 0.00324 0.00324 0.00324 0.00324 0.00324 0.00324 Depletion Uncertainty 0.00217 0.00441 0.00746 0.00876 0.00992 0.01116 0.01267 Burnup Uncertainty 0.00255 0.00479 0.00601 0.00618 0.00703 0.00793 0.00932 Statistical Combination of Uncertainties 0.01281 0.01397 0.01564 0.01637 0.01735 0.01845 0.02001 Total Biases and Uncertainties 0.03040 0.03293 0.03185 0.03274 0.03351 0.03386 0.03687 Targeted kmax (95/95) 0.99500 0.99500 0.99500 0.99500 0.99500 0.99500 0.99500 Targeted keff 0.96460 0.96207 0.96315 0.96226 0.96149 0.96114 0.95813 Interpolated Burnup (GWd/MTU) 0.11 7.41 16.74 21.18 28.05 33.69 38.27 Polynomial Burnup (GWd/MTU) 4.08 11.57 18.42 24.71 30.49 35.84 40.83 1BTable 4 Loading Curve for Region 2 Loading Configurations Minimum Required Fuel Assembly Burnup (GWd/MTU) as a Function of Initial Enrichment (wt% U-235)

Uniform Loading for Region 2 f(x) = 0.0894x3 - 1.9403x2 + 22.344x - 33.56 (Note 1)

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

ATTACHMENT 1 Evaluation of the Proposed Change Page 11 of 15 Figure 2 Loading Curve for Region 2 The same normal operation scenarios as the original submittal (Ref. 4.1.1) were considered.

The same abnormal and accident scenarios were evaluated as described in the original submittal (Ref. 4.1.1) for enrichments up to 5.0 wt%, except for the mislocation event for the "G" spent fuel pool rack at Braidwood. A description of the "G" spent fuel pool rack at Braidwood is provided in the CEG response to Question 4 of Attachment 2.

As a bounding approach, the mislocation of a single fresh fuel assembly with an enrichment of 5.0 wt% in a rack corner facing two adjacent fuel assemblies was considered. The attached Boral panels of all exterior storage rack walls at the rack corner as well as all cell blockers were not credited (see Figure 3). 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.

50 45 40

~35 -

-0

$ 30

~

-;: 25

J C

~ 20 ca (I) 15

J LL 10 5

0 2

Minimum Required Fuel Assembly Burn up as a Function of Nominal Initial Enrichment 2.5 Loading Curve ACCEPTABLE BURNUP DOMAIN UNACCEPTABLE BURNUP DOMAIN 3

~5 4

Initial Fuel Enrichment (wt% U-235) 4.5 5

ATTACHMENT 1 Evaluation of the Proposed Change Page 12 of 15 Figure 3 Radial Cross Section View of MCNP Model for Mislocated Fuel Assembly The bounding results for mislocated fuel assembly, single assembly misload, and multiple assembly misload accidents are summarized in Table 5. Table 5 supersedes the information previously provided in Ref. 4.2.1, Attachments 1 and 8 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.

Mislocated Fuel Assembly

ATTACHMENT 1 Evaluation of the Proposed Change Page 13 of 15 2BTable 5 Results of Accident Cases Description Soluble boron Concentration (ppm) keff Uncertainty kmax (95/95)

(Note 1)

Mislocated Fuel Assembly Cases, 5.0 wt%, 40.82 GWd/MTU Cell centered 0

1.03754 0.00038 1.07441 Toward the mislocated fuel assembly 0

1.03396 0.00038 1.07083 Cell centered 1000 0.90888 0.00037 0.94575 Toward the mislocated fuel assembly 1000 0.91058 0.00037 0.94745 Required Soluble Boron Concentration (ppm) 979.9 Single Assembly Misload Cases, 5.0 wt%, 40.82 GWd/MTU Cell centered 0

0.96325 0.00038 1.00012 Rack centered 0

0.97220 0.00035 1.00907 Cell centered 1000 0.86780 0.00033 0.90467 Rack centered 1000 0.87606 0.00036 0.91293 Required Soluble Boron Concentration (ppm) 614.4 Multiple Assembly Misload Cases, 5.0 wt%, 40.82 GWd/MTU Cell centered 0

1.00694 0.00037 1.04381 Rack centered 0

1.01419 0.00037 1.05106 Cell centered 1000 0.90952 0.00036 0.94639 Rack centered 1000 0.91580 0.00037 0.95267 Cell centered 2000 0.83467 0.00035 0.87154 Rack centered 2000 0.84006 0.00036 0.87693 Required Soluble Boron Concentration (ppm) 1116.7 Note 1: The applicable Total Biases and Uncertainties determined in Table 3 is used to determine kmax (95/95).

Evaluation of the interface conditions with Region 2 storage racks remain the same as described in the original submittal (Ref. 4.1.1). Updated results for the Region 1 to Region 2 interface, based on an enrichment up to 5.0 wt% and the changes to the applicable biases and uncertainties described above, are summarized in Table 6. Table 6 supersedes the information previously provided in Ref. 4.2.1, Attachments 1 and 8 Table 11. The reactivity of the interface is confirmed to remain below the regulatory limit.

ATTACHMENT 1 Evaluation of the Proposed Change Page 14 of 15 3BTable 6 Results of Region 1 to Region 2 Interface Cases Description Soluble boron Concentration (ppm) keff Uncertainty kmax (95/95)

(Note 1) 2.0 wt%, 4.07 GWd/MTU All Fuel Assemblies Cell Centered 0

0.92367 0.00032 0.96054 Fuel Assemblies Toward to Interface Center 0

0.91992 0.00033 0.95679 All Fuel Assemblies Cell Centered 500 0.86330 0.00042 0.90017 Fuel Assemblies Toward to Interface Center 500 0.86157 0.00042 0.89844 3.5 wt%, 24.70 GWd/MTU All Fuel Assemblies Cell Centered 0

0.93664 0.00035 0.97351 Fuel Assemblies Toward to Interface Center 0

0.93432 0.00035 0.97119 All Fuel Assemblies Cell Centered 500 0.87833 0.00035 0.91520 Fuel Assemblies Toward to Interface Center 500 0.87442 0.00037 0.91129 5.0 wt%, 40.82 GWd/MTU All Fuel Assemblies Cell Centered 0

0.93492 0.00037 0.97179 Fuel Assemblies Toward to Interface Center 0

0.93438 0.00035 0.97125 All Fuel Assemblies Cell Centered 500 0.88406 0.00036 0.92093 Fuel Assemblies Toward to Interface Center 500 0.88244 0.00035 0.91931 Note 1: The applicable Total Biases and Uncertainties determined in Table 3 is used to determine kmax (95/95).

ATTACHMENT 1 Evaluation of the Proposed Change Page 15 of 15 4.

REFERENCES 4.1.

Licensing Evaluation References 4.1.1. Letter from K. Lueshen (Constellation Energy Generation, LLC) to U.S. NRC, "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", dated September 29, 2023 (ADAMS Accession No. ML23272A201) 4.1.2. 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) 4.2.

Benchmarking / Criticality Analyses References 4.2.1. BRW-24-0031-N/BYR24-011, Revision 0, "Braidwood-Byron GAIA Criticality Analysis for New Fuel Vault", dated October 2024 4.2.2. BRW-24-0032-N/BYR24-012 Revision 0, "Braidwood-Byron GAIA Criticality Analysis for Spent Fuel Pool", dated October 2024

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 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 Response to Request for Additional Information (Non-Proprietary)

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 1 of 36 REQUEST FOR ADDITIONAL INFORMATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION BRAIDWOOD, UNITS 1 AND 2, AND BYRON, UNITS 1 AND 2, TS 3.7.15, SFP BORON, 3.7.16, SFA STORAGE, 4.3.1 FUEL STORAGE, CRITICALITY CONSTELLATION ENERGY GENERATION, LLC BRAIDWOOD, UNITS 1, 2; BYRON, UNITS 1, 2 DOCKET NO. 05000454, 05000455, 05000456, 05000457

Background

By letter dated September 29, 2023, Agencywide Documents Access and Management System (ADAMS) Accession No. ML23272A201) Constellation Energy Generation, LLC (Constellation or the licensee) submitted a license amendment (LAR) request to change the Braidwood Station (Braidwood), Units 1 and 2, and Byron Station (Byron), Unit Nos. 1 and 2, Technical Specifications (TSs) 3.7.15, "Spent Fuel Pool Boron Concentration", 3.7.16, "Spent Fuel Assembly Storage", and TS 4.3.1 "Fuel Storage, Criticality". A supporting criticality safety evaluation (SE) for fuel assembly storage in the Braidwood and Byron spent fuel pool (SFP) storage racks and new fuel storage vaults (NFV). The criticality SE was performed to support a fuel transition to Framatome GAIA fuel assemblies and potential future transition to 24-month cycles, Framatome GAIA AFM fuel assemblies, and increasing the fuel assembly maximum uranium-235 (U-235) enrichment from 5.0 weight percent (%) (wt/%) to 6.5 wt/%. However, the LAR did not request a change to Braidwood and Byron TS 4.3.1.a to raise the maximum licensed U-235 enrichment nor did the LAR request an exemption to Title 10 of the Code of Federal Regulations (10 CFR) 50.68(b)(7) which limits the maximum U-235 enrichment to 5.0 wt/%.

The proposed amendment is to change:

x TS 3.7.15, "Spent Fuel Pool Boron Concentration" to increase the required spent fuel pool boron concentration to be > 2000 part per million (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 percent probability, 95 percent 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 percent probability, 95 percent 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)

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 2 of 36 Regulatory Basis In accordance with the licensees amendment request, the regulatory requirements and guidance which the U.S. Nuclear Regulatory Commission (NRC) staff considered in assessing the proposed TS change are as follows:

Pursuant to 10 CFR, part 50, appendix A, criterion 62, Prevention of criticality in fuel storage and handling, requires that criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

Pursuant to 10 CFR 50.68(a), each holder of an operating license shall comply with either 10 CFR 70.24 or the requirements in 10 CFR 50.68(b). The licensee has elected to meet 10 CFR 50.68(b). Accordingly, and as relevant to this license amendment request, the licensee must comply with the following 50.68(b) requirements:

(1) Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.

(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.

(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.

(4) If no credit for soluble boron is taken, 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 categories of items required to be in the TSs are provided in 10 CFR 50.36(c).

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 3 of 36 Question 1 Constellation's letter dated September 29, 2024, the LAR proposed replacing TS Figure 3.7.16--1. The proposed TS Figure 3.7.16-1, Minimum Required Fuel Assembly Burnup as a Function of Nominal Initial Enrichment, specifies the storage requirements for the SFP, Region 2. The proposed TS Figure 3.7.161 does not have any consideration for decay time. Yet the LAR ((

)) As currently proposed the licensee would be in violation of their proposed TS ((

)). Therefore, the NRC staff requests the licensee propose a TS that accommodates ((

)), or discuss how they intend to bring the SFP into compliance with the proposed TS, given the above discussion regarding ((

)).

Constellation Response to Question 1 In the prior submittal, it was noted that the assembly sustained grid damage and could not be moved. However, it was confirmed that the grid damage does not impact the handling of the assembly and as such, can be moved with extra controls, challenges, and oversight.

Constellation will [

]C. Constellation will be moving the assembly in question into Region 1 storage racks prior to LAR approval. This is being tracked using a regulatory commitment, similar to the previously submitted Braidwood commitment. RS-24-098 Attachment 9 lists both the original Braidwood commitment and the new Byron commitment for ease of tracking.

Question 2 The LAR states in section 3.1, for all proposed revised results, the guidance from Nuclear Energy Institute (NEI) 12-16, "Guidance for Performing Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," and Regulatory Guide (RG) 1.240, "Fresh and Spent Fuel Pool Criticality Analyses," (References. 6.1.4 and 6.1.5), has been applied. RG 1.240, paragraph C.1.O, states, "NEI 12-16, Revision 4, provides many recommendations that are based on analyses performed using typical geometries and compositions associated with SFPs and bundle designs that are currently in widespread use in the United States (U.S.) (e.g.,

cylindrical uranium dioxide fuel pellets enclosed in zirconium alloy tubes). Novel configurations and concepts, such as accident-tolerant fuel designs, may require justification for continued use of the assumptions. For example, dispositions of specific uncertainties as not significant may no longer be valid, simplifying assumptions may become nonconservative, and additional uncertainties may need to be considered". Fuel enrichments above 5.0 wt/% U-235 are outside the typical geometries and compositions associated with SFPs and bundle designs that were in widespread use in the United U.S. when the guidance in NEI 12-16 and RG 1.240 was developed. However, the licensees LAR does not provide any information that justifies the application of RG 1.240 and NEI 12-16 guidance to its analysis. The NRC requests the licensee provide the justification for applying RG 1.240 and NEI 12-16 guidance to its analysis.

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 4 of 36 Constellation Response to Question 2 Constellation has performed analyses which support loading GAIA fuel with enrichment limited to 5.0 wt% into the new fuel vault and spent fuel pool. These analyses are documented in BRW-24-0031-N/BYR24-011 and BRW-24-0032-N/BYR24-012 (Refs. [1] and [2]), respectively.

Therefore, this question no longer needs to be answered as CEG is not pursuing greater than 5.0 wt% U-235 fuel.

Question 3 The LAR used CASMO5 as the depletion code for the burnup credit portion of the analysis, including depletion of fuel with an initial U235 enrichment up to 6.5 wt/%. Per NRCs approval of CASMO5 in Studsvik topical report SSP-14-P01/028-TR-P-A, Generic Application of the Studsvik Scandpower Core Management System to Pressurized Water Reactors, Revision 0 (ADAMS ML17279A985) CASMO5 is limited to U-235 enrichments of 5.0 wt/% U-235 and below. In order to address the apparent discrepancy between the NRC approved scope for CASMO5 and the proposed applicability within the LAR, clarify if Constellation requests to obtain NRC review and approval for the SFP criticality analysis and corresponding proposed TS curves for enrichments up to 6.5 wt/%. If this was not the intent, then provide updates to the LAR submittal to clarify that the scope of the request is limited to enrichments up to 5.0 wt/%,

including corresponding updates to the proposed TS curves and SFP soluble boron requirements. If approval for the criticality analysis is requested for up to 6.5 wt/%, then provide justification that CASMO5, as used by the licensee in support of this LAR, is acceptable for the extended enrichment range in a manner that is consistent with how CASMO5 was qualified for enrichments up to 5.0 wt/%.

Constellation Response to Question 3 Constellation has performed analyses which support loading GAIA fuel with enrichment limited to 5.0 wt% into the new fuel vault and spent fuel pool. These analyses are documented in BRW-24-0031-N/BYR24-011 and BRW-24-0032-N/BYR24-012 (Refs. [1] and [2]), respectively.

Therefore, this question no longer needs to be answered as CEG is not pursuing greater than 5.0 wt% U-235 fuel.

Question 4 The LAR did not provide specifics with regard to its SFP rack model design. During a virtual audit, the NRC staff reviewed details of the licensees SFP rack modules. The NRC staff noted that the Braidwood SFP, module G, was modified or repaired. The NRC staff requests the following information to complete its review:

x Information related to the as built documentation for the SFPs and the models used in the analysis.

x Information related to the modified/repaired Braidwood SFP, module G. Include as-left dimensions and materials, impact on calculation of keff, impact on accident analysis, and seismic qualification.

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 5 of 36 Constellation Response to Question 4 Region 1:

Fuel assemblies are placed in Region 1 such that a nominal 10.888 inch north-south and 10.574 inch east-west, center-to-center distance is maintained between fuel assemblies. A total of 396 storage cells are provided in four racks in Region 1. The cells provide a smooth and continuous surface for lateral contact with the fuel assembly. The rack module components, as related to this analysis, are as follows:

a.

Internal Square Box: This element provides the lateral bearing surface to the fuel assembly. It is fabricated by joining two formed channels using a controlled seam welding operation. This element has an 8.82-inch square internal cross-section and is 167 inches long.

b.

Neutron Absorber Material (Boral): Boral is placed on all four sides of a square tube over a length of 147 inches.

c.

Poison Sheathing: The poison sheathing (cover plate) serves to position and retain the poison material in its designated space. This is accomplished by spot welding the cover sheet to the square tube along the former's edges at numerous (at least 19) locations.

This manner of attachment ensures that the poison material will not sag or laterally displace during fabrication processes and under any subsequent loading conditions.

d.

Gap Element: Gap elements position two inner boxes at a predetermined distance to maintain the minimum flux trap gap required between two boxes. The gap element is welded to the inner box by fillet welds. An array of composite box assemblies forms the honeycomb gridwork of cells which harnesses the structural strength of all sheet and plate type members in an efficient manner. The array of composite boxes has overall bending, torsional, and axial rigidities which are an order of magnitude greater than configurations utilizing a grid bar type of construction.

The nominal values presented in Table 4-1 below 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, with no credit for fuel burnable absorber rods (such as Gd2O3).

Reflective boundary conditions were used on all 6 sides, making the models infinite. Figure 4-1 shows the radial cross section views of the modeled fuel assembly and rack.

As indicated in Table 4-1, the distance between two Region 1 racks is more than the flux trap between two Region 1 cells. Thus, conservatively, the distance between two Region 1 racks is reduced to the flux trap.

The sites coupon surveillance procedures accept [

]C, conservatively, [

]C Boron-10 loss was considered as the Boral B-10 degradation bias in this report. This was conservatively done by reducing the Boral density.

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 6 of 36 A sensitivity analysis was also performed to evaluate the effect of the Boral with the minimum density (instead of the minimum B-10 areal density) on keff, to demonstrate that the modeling of the Boral with the minimum areal density was conservative.

The following rack manufacturing uncertainties were evaluated:

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

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 7 of 36 Table 4-1: SFP Regions 1 Specifications Specification Value Region 1 Cell inside width (in.)

8.82 +/- [

]C Steel wall thickness (in.)

[

]C Boral absorber thickness (in.)

[

]C Boral absorber width (in.)

[

]C Boral absorber length (in.)

(Note 1) 147 + [

]C Boral minimum B-10 areal density (g/cm2)

[

]C Steel sheath thickness (in.)

[

]C N-S lattice spacing (in.)

10.888 +/- [

]C E-W lattice spacing (in.)

10.574 +/- [

]C N-S flux trap (in.)

[

]C E-W flux trap (in.)

[

]C Minimum gap between Region 1 Racks (in.)

[

]C Note 1. Conservatively, it is assumed that the boral absorber length is the same as active fuel length.

Figure 4-1 Radial Cross Section View of MCNP Model of the Base Case Fuel Assembly in the Region 1 Rack Region 2:

The rack data of Region 2, including Boral composition and density, are provided in Table 4-2.

Fuel assemblies are placed in Region 2 such that a nominal 8.97 inch center-to-center distance is maintained between fuel assemblies. A total of 2588/2568 storage cells are provided in Region 2 at Byron/Braidwood respectively. The cells provide a smooth and continuous surface

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 8 of 36 for lateral contact with the fuel assembly. The rack module components, as related to this analysis, are as follows:

a.

Internal Square Box: This element provides the lateral bearing surface to the fuel assembly. Stainless steel boxes are arranged in an alternating pattern such that the connection of the box corners form storage cells between those of the stainless steel boxes. This element has an 8.75-inch square internal cross-section and is 167 inches long.

b.

Neutron Absorber Material (Boral): Boral panel is attached by a stainless steel sheathing centered on each side of a square box over a length of 147 inches.

c.

Poison Sheathing The poison sheathing (cover plate) serves to position and retain the poison material in its designated space. This is accomplished by spot welding the cover sheet to the square tube along the former's edges at numerous locations. This manner of attachment ensures that the poison material will not sag or laterally displace during fabrication processes and under any subsequent loading conditions.

The reactivity effect of a decrease of [

]C in Boron-10 is considered as the Boral B-10 degradation bias in this analysis.

The nominal values presented in Table 4-2 were used to build the MCNP model of the base case Region 2 racks.

The base case model included a 2x2 array Region 2 cells. Axially, only the active region of the fuel assembly was modeled. 30 cm water reflector were 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 Region 2 racks was neglected. Figure 4-2 show the radial cross section views of the modeled fuel assembly and rack.

The Boral was modelled with the minimum areal density which was conservative based on NEI 12-16 (Ref. [9]).

The following rack manufacturing uncertainties were evaluated:

x Box ID x

Box wall thickness x

Poison width x

Poison thickness x

Cell pitch

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 9 of 36 Table 4-2: SFP Regions 2 Specifications Specification Value Region 2 Cell inside width (in.)

8.75 +/- [

]C Steel wall thickness (in.)

[

]C Boral channel thickness (in.)

[

]C Boral absorber thickness (in.)

[

]C Boral absorber width (in.)

[

]C Boral absorber length (in.)

(Note 1) 147 [

]C Boral minimum B-10 areal density (g/cm2)

[

]C Steel sheath thickness (in.)

[

]C Lattice spacing (in.)

8.97 +/- [

]C Minimum gap between Region 2 Racks (in.)

[

]C Minimum gap between Region 1 and Region 2 Racks (in.)

[

]C Note 1. Conservatively, it is assumed that the boral absorber length is the same as active fuel length.

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 10 of 36 Figure 4-2 Radial Cross Section View of MCNP Model of the Base Case Fuel Assembly in Region 2 Racks Braidwood Rack G Modification:

The area of "G" spent fuel pool rack directly beneath the return pipe at Braidwood was found damaged and the damaged rack pieces were removed from the pool in 2008, creating an open area atop rack base plate without any storage locations. Boral panels and sheathing around this open area were removed from the "G" spent fuel pool rack, so that local reinforcing bars were added at the top and bottom ends of the rack adjacent cell walls to stabilize the structure. Cell block devices were also installed into 8 storage locations around the open area to prevent insertion of full assemblies in those cells. The modified "G" rack was placed back into the spent fuel pool. The locations of the open area in the "G" spent fuel pool rack are shown in Figure 4-3.

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ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 11 of 36 Figure 4-3 Locations of Open Areas in the "G" Spent Fuel Pool Rack As shown in Figure 4-3, an accident condition could possibly occur when a fuel assembly is accidentally mislocated in the open area of the "G" spent fuel pool rack. However, the mislocated fuel assembly would either be located in a rack corner facing two adjacent cell block devices or face at most one adjacent fuel assembly in one direction. As a bounding approach, calculations were performed with a single fresh fuel assembly of the highest permissible enrichment (5.0 wt% 235U) mislocated in a rack corner facing two adjacent fuel assemblies.

Boral neutron absorber panels attached at the exterior storage rack walls around the mislocated fuel assembly were not considered, and conservatively, all cell blockers were not credited, as shown in Figure 3 of RS-24-098 Attachment 1. Both the cell centered and eccentric fuel positioning were considered. The minimum soluble boron concentrations were determined by linear interpolation to ensure that the kmax(95/95) value does not exceed the regulatory limit of 0.95. The results summarized in Table 4-3 indicate that it is bounded by the required soluble boron concentration of 2000 ppm from the analysis with sufficient margin.

The modified "G" spent fuel rack does not have impact on spent fuel pool seismic qualification and the existing seismic qualification of the spent fuel rack will not be adversely impacted the storage of the GAIA fuel assemblies.

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ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 12 of 36 Table 4-3: Results of Mislocated Fuel Assembly Cases Description Soluble boron Concentration (ppm) keff Uncertainty kmax (95/95)

(Note 1) 2.0 wt%, 4.07 GWd/MTU Cell centered 0

1.04909 0.00037 1.08596 Toward the mislocated fuel assembly 0

1.04691 0.00037 1.08378 Cell centered 1000 0.89918 0.00035 0.93605 Toward the mislocated fuel assembly 1000 0.90191 0.00036 0.93878 Required Soluble Boron Concentration (ppm) 923.8 3.5 wt%, 24.70 GWd/MTU Cell centered 0

1.04065 0.00040 1.07752 Toward the mislocated fuel assembly 0

1.03850 0.00037 1.07537 Cell centered 1000 0.90602 0.00038 0.94289 Toward the mislocated fuel assembly 1000 0.90747 0.00037 0.94434 Required Soluble Boron Concentration (ppm) 957.5 5.0 wt%, 40.82 GWd/MTU Cell centered 0

1.03754 0.00038 1.07441 Toward the mislocated fuel assembly 0

1.03396 0.00038 1.07083 Cell centered 1000 0.90888 0.00037 0.94575 Toward the mislocated fuel assembly 1000 0.91058 0.00037 0.94745 Required Soluble Boron Concentration (ppm) 979.9 Note 1: The applicable Total Biases and Uncertainties determined in Table 8.2.15 of [2] is used to determine kmax (95/95).

Question 5 The LAR states that Boral is used as the SFP neutron absorbing material (NAM). In its LAR, Constellation states that degradation of Boral is treated as an uncertainty. Treating degradation of SFP NAM as an uncertainty is inconsistent with the current guidance in NEI 12-16 R4 and appears to be non-conservative. The LAR does not identify how treating Boral degradation as an uncertainty impacts the calculation of keff.

During audit discussions, Constellation stated that NEI 12-16 R4, section 5.2.2, indicates neutron absorber material as-built dimensions should be treated as uncertainties. While that is correct, Constellation states in its September 29, 2023, letter that Poison thickness (resulting in a boron areal density less than the minimum) is listed as an uncertainty. The guidance in NEI 12-16 R4, section 5.2.2, for treating neutron absorber material manufacturing thickness tolerance as an uncertainty is only with regard to the thickness of the material and not the boron

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 13 of 36 areal density of the neutron absorber material. Any reduction of the minimum boron areal density must be treated as a bias.

During audit discussions, the licensee quoted NEI 16-03 R1, to date no degradation trend leading to loss of 10B has been observed for BORAL. And went on to state, For this reason, no degradation is anticipated for the Boral as neutron absorber in Byron and Braidwood SFP racks. However, the guidance in NEI 12-16 R4, section 5.2.2.3.2, states, if degradation is anticipated to result in loss of 10B areal density or absorber effectiveness [emphasis added],

then appropriate margin to account for the degradation needs to be included in the criticality analysis Boral has a propensity to form blisters, the void filled blisters will reduce the absorber effectiveness.

To continue its review, the NRC staff requests the licensee provide the following information:

Detailed description of how the Boral neutron absorbing materials were modeled in the base cases versus the design and the perturbations made to the models to determine the impact on reactivity.

Discussion of Boral blistering at Byron and Braidwood including the extent of blistering and when blistering would affect the criticality analysis.

Constellation Response to Question 5 The Boral B-10 degradation is treated now as a bias.

The Braidwood and Byron coupon procedures state, [

]C While the procedures accept [

]C, conservatively, this

[

]C was considered as the Boral B-10 degradation bias in this analysis. This was conservatively done by reducing the Boral density from [

]C.

Consequently, the Boral B-10 areal density was also decreased from [

]C. No other changes were made to the model.

To date, no blistering has been observed in Byron or Braidwood Boral coupons.

Question 6 In its LAR, Constellation stated, The criticality safety analysis of record (Reference 6.1.9)

[HI-982094, Revision 5, "Criticality Evaluation for the Byron/Braidwood Rack Installation Project", dated December 2013] 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). The Byron and Braidwood TS 4.3.1.b identifies Holtec International Report HI-982094, Criticality Evaluation for the Byron/Braidwood Rack Installation Project, Project 80944 1998 as the analysis of record.

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 14 of 36 To continue its review the NRC staff requests the licensee provide the following information:

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

Constellation Response to Question 6 Non-proprietary and proprietary versions of the requested evaluation are provided in RS-24-098 Attachments 10 and 11, respectively. Note that the provided documents are Revision 6. Revision 5 of the Holtec report was issued in 2013 with only a proprietary version.

To facilitate response to this question without requiring withholding of the entire document, Revision 6 was created with the only change being the administrative addition of brackets to designate proprietary information contained within the report.

The single page before of the Holtec report is a plant-specific evaluation for impact based on Information Notice 2011-03. The page reflects Revision 5 of the Exelon/Constellation Design Analysis, as that remains the analysis of record for Westinghouse fuel at Braidwood and Byron.

Revision 6 of the Holtec report was added into Braidwood and Byron records management as a minor revision to major Revision 5.

Question 7 In its LAR, Constellation included a fuel depletion related geometry change bias. The LAR is not clear whether GAIA or GAIA AFM fuel designs were used, which enrichment was used, and whether the current or future 24-month cycle length was used. Justification for what was done in the analysis is not clearly demonstrated in the LAR. In order for the NRC staff to complete its review, it needs to understand how the calculation of the fuel depletion related geometry change bias impacts the calculation of keff.

Therefore, the NRC staff requests the following information:

Whether GAIA or GAIA AFM fuel designs were used.

Which enrichment was used.

What assumptions were used for fuel residence time (i.e., number of cycles and cycle duration).

Provide justification for what was done in the analysis.

Explain how the calculation of the fuel depletion related geometry change bias impacts the calculation of keff.

Constellation Response to Question 7 As CEG is not pursuing greater than 5.0 wt% U-235 fuel, only GAIA fuel designs are used in the criticality analyses for enrichments up to 5.0 wt% and 18-month cycles.

The important fuel depletion related changes considered for the analysis are fuel rod growth and cladding creep, and the material dependent grid growth. These fuel assembly geometry changes can affect the neutron spectrum during depletion by changing the fuel to moderator ratio and isotopic composition. The effects of these possible fuel geometry changes on the reactivity of the fuel in the SFP are discussed below.

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Page 15 of 36 x

Fuel rod growth and cladding creep: Fuel rod growth and cladding creep have the potential to decrease the fuel to moderator ratio in the geometry, thus potentially increasing reactivity. The creep model considers the pellet to clad gap essentially closed (pellet-to-clad gap of [

]C inches) and the clad thinning based on the maximum fuel rod growth after depletion of 1.44 inches. The difference between the keff of the cases with fuel rod growth and the base case keff was considered as bias, if positive.

x Grid growth: Grid growth increases the pitch between fuel rods thus decreases the fuel to moderator ratio, consequently increasing reactivity. The maximum grid growth was based on the tolerance of rod pitch of [

]C inches. It was conservatively assumed that all fuel rods in the fuel assembly have the maximum rod pitch due to grid growth.

The difference between the keff of the cases with grid growth and the base case keff was considered as bias, if positive.

Separate depletion calculations were performed with a single full power cycle to achieve the desired burnup for the two above cases. The maximum positive reactivity effects of two cases above were added together as bias and the 95/95 uncertainty of the bias was statistically combined with other uncertainties. The results are shown in Table 7-1 below.

Table 7-1: Results for Fuel Assembly Geometry Change with Depletion Cases Description keff Uncertainty Delta-k 95/95 unc 2.0 wt%, 5 GWd/MTU Base Case 0.92306 0.00030 Fuel rod growth and cladding creep 0.92537 0.00031 0.00231 0.00086 Grid spacer growth 0.92500 0.00031 0.00194 0.00086 Total Bias/Statistical Combination of Uncertainties 0.00425 0.00122 3.5 wt%, 20 GWd/MTU Base Case 0.96825 0.00032 Fuel rod growth and cladding creep 0.97055 0.00035 0.00230 0.00095 Grid spacer growth 0.97141 0.00034 0.00316 0.00093 Total Bias/Statistical Combination of Uncertainties 0.00546 0.00133 5.0 wt%, 35 GWd/MTU Base Case 0.98023 0.00034 Fuel rod growth and cladding creep 0.98287 0.00034 0.00264 0.00096 Grid spacer growth 0.98374 0.00036 0.00351 0.00099 Total Bias/Statistical Combination of Uncertainties 0.00615 0.00138 Question 8 In its LAR, Constellation did not identify the nuclides that were modeled in its SFP criticality analysis. NEI 12-16 R4 states, One important consideration is that certain nuclides, such as fission gases and short-lived nuclides will no longer be present in the fuel matrix to the extent predicted by the depletion code. For the NRC staff to complete its review it requires

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Page 16 of 36 information regarding which nuclides were included in the analysis.

Therefore, the NRC staff requests the following information:

The nuclides that were modeled in the SFP criticality analysis.

Justification for those nuclides that were modeled.

Details regarding how the fission gas release bias used in the analysis was calculated.

Constellation Response to Question 8 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.

A list of credited spent fuel isotopes was shown in Table 8-1 below. Note that Xe-135 is not considered in the list of isotopes.

Table 8-1: List of Spent Fuel Isotopes Used in Analysis U-235 U-234 U-238 O-16 O-17 Th-229 Th-230 Th-231 Th-232 Th-233 Th-234 Pa-231 Pa-232 Pa-233 U-232 U-233 U-236 U-237 U-239 U-240 Np-234 Np-235 Np-236 Np-237 Np-238 Np-239 Pu-236 Pu-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Pu-242 Pu-243 Am-241 Am-242 Am-242m Am-243 Am-244m Am-244 Cm-242 Cm-243 Cm-244 Cm-245 Cm-246 Cm-247 Cm-248 Cm-249 Cm-250 Bk-245 Bk-246 Bk-247 Bk-248 Bk-249 Bk-250 Cf-248 Cf-249 Cf-250 Cf-251 Cf-252 Ge-76 As-74 As-75 Se-76 Se-77 Se-78 Se-79 Se-80 Se-82 Br-79 Br-81 Kr-80 Kr-82 Kr-83 Kr-84 Kr-85 Kr-86 Rb-85 Rb-86 Rb-87 Sr-88 Sr-89 Sr-90 Y-89 Y-90 Y-91 Zr-90 Zr-91 Zr-92 Zr-93 Zr-94 Zr-95 Zr-96 Nb-93 Nb-94 Nb-95 Mo-94 Mo-95 Mo-96 Mo-97 Mo-98 Mo-99 T-99 Ru-98 Ru-99 Ru-100 Mo-100 Ru-101 Ru-102 Ru-103 Ru-104 Ru-105 Ru-106 Rh-103 Rh-105 Pd-104 Pd-105 Pd-106 Pd-107 Pd-108 Ag-107 Ag-109 Ag-110m Cd-106 Cd-108 Cd-110 Pd-110 Ag-111 Cd-111 Cd-112 Cd-113 Cd-114 Cd-116 In-113 In-115 Sn-112 Sn-113 Sn-114 Sn-115 Sn-116 Sn-117 Sn-118 Sn-119 Sn-120 Sb-121 Sn-122 Sn-123 Sn-124 Sn-125 Sn-126 Sb-123 Sb-124 Sb-125 Sb-126 Te-122 Te-123 Te-124 Te-125 Te-126 Te-127m Te-128 Te-129m Te-130 Te-132 I-127 I-129 I-130 I-131 Xe-128 Xe-129 Xe-130 Xe-131 Xe-132

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Page 17 of 36 I-135 Xe-133 Xe-134 Xe-136 Cs-133 Cs-134 Cs-135 Cs-136 Cs-137 Ba-134 Ba-135 Ba-136 Ba-137 Ba-138 Ba-140 La-138 La-139 La-140 Ce-138 Ce-139 Ce-140 Ce-141 Ce-142 Ce-143 Ce-144 Pr-141 Pr-142 Pr-143 Nd-142 Nd-143 Nd-144 Nd-145 Nd-146 Nd-147 Nd-148 Nd-150 Pm-147 Pm-148m Pm-148 Pm-149 Pm-151 Sm-147 Sm-148 Sm-149 Sm-150 Sm-151 Sm-152 Sm-153 Sm-154 Eu-153 Eu-154 Eu-155 Eu-156 Eu-157 Gd-153 Gd-154 Gd-155 Gd-156 Gd-157 Gd-158 Gd-160 Tb-159 Tb-160 Dy-160 Dy-161 Dy-162 Dy-163 Dy-164 Several nuclides are credited in the depletion and criticality analysis base cases, including major actinides, minor actinides, and fission products. However, certain nuclides, such as fission gases nuclides, may no longer be present in the fuel matrix and fuel rods. Conservative gas release fractions have been reviewed and approved by the NRC in Regulatory Guide 1.183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (Ref. [7]). The fission gas release fraction of each radionuclide group was therefore determined based on the values provided in Table 3 of RG 1.183.

Consequently, the reactivity difference between the keff of the cases with the fraction of fission gases removed and the base case keff was considered as the bias of fission gas release, and the 95/95 uncertainty of the bias was statistically combined with other uncertainties. The results are shown in Table 8-2 below.

Table 8-2: Results for Fission Gas Release Cases Description Enrichment (wt.%)

Burnup (GWd/MTU) keff Uncertainty Delta-k 95/95 unc Base Case 2

5 0.92306 0.00030 Fission Gas Release 0.92330 0.00033 0.00024 0.00089 Base Case 2.5 10 0.94156 0.00033 Fission Gas Release 0.94231 0.00034 0.00075 0.00095 Base Case 3

20 0.93100 0.00033 Fission Gas Release 0.93210 0.00036 0.00110 0.00098 Base Case 3.5 25 0.94292 0.00034 Fission Gas Release 0.94418 0.00035 0.00126 0.00098 Base Case 4

30 0.95222 0.00036 Fission Gas Release 0.95327 0.00037 0.00105 0.00103 Base Case 4.5 35 0.95401 0.00034 Fission Gas Release 0.95431 0.00034 0.00030 0.00096

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Page 18 of 36 Base Case 5

40 0.94643 0.00036 Fission Gas Release 0.94818 0.00036 0.00175 0.00102 Question 9 In its LAR, Constellation states, Depletion calculations were performed with conservative operating conditions: highest fuel temperature, moderator temperature and soluble boron concentrations during in-core operation. For the NRC staff to complete its review it needs to understand which depletion parameters were used in the depletion portion of the analysis and the impact on the calculation of keff.

Therefore, the NRC staff requests the following information:

Provide details about what was used in the depletion calculation.

Provide a comparison to what was used to actual past operating conditions.

Demonstrate how the depletion calculation bounds future operations with the fuel changes, Demonstrate how the depletion calculation bounds the transition to 24-month cycles, Demonstrate how the depletion calculation bounds U235 enrichment up to 6.5 wt/%

Provide a justification for the use of any depletion parameters that are not bounding.

Describe the difference between bounding parameters and those used in the analysis.

Constellation Response to Question 9 As CEG is no longer pursuing greater than 5.0 wt% U-235 fuel; therefore, neither 24-month cycles nor enrichments up to 6.5 wt% need to be considered.

Studies performed by Oak Ridge National Laboratory have identified that higher moderator and fuel temperatures and higher soluble boron concentration during depletion result in increased reactivity of used fuel in the storage rack, while the power density has a much smaller effect on reactivity than the impact of moderator and fuel (Refs. [3] and [4]). Also, a lower power density would be inconsistent with the higher fuel temperature. Consistent with the NEI 12-16 (Ref. [9]),

the upper bound core operating parameters from the input data are conservatively considered for Region 2 analysis and used for fuel depletion calculations performed with CASMO5.

Core operating parameters are presented in Table 9-1 below. The comparison of the used moderator temperature with the values from other references is shown in Tables 9-2. The comparison of the used fuel temperature with the values from other references is shown in Tables 9-3. Specifically, the moderator temperature considered in the analysis bounds the highest core outlet temperature and hot leg temperature, and the fuel temperature considered in the analysis is much higher than the core average fuel temperature.

Table 9-1: Core Operating Parameters Specification Value Maximum Core Moderator Temperature (K) 600 Maximum Fuel Temperature (K) 1050 Reactor Specific Power (W/gU) 39.69 Soluble Boron Concentration (ppm) 1000

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Page 19 of 36 Specification Value In-Core Assembly Pitch (in.)

8.466 Table 9-2: Comparison of Moderator Temperature Specification Value K

F Moderator Temperature Used in the Analysis 600 620.33 Maximum Hot Leg Temperature Byron Unit 1 617 Byron Unit 2 612 Braidwood Unit 1 617.29 Braidwood Unit 2 610.04 Core Outlet Temperature Byron Unit 1 Cycle 26 619.52 Byron Unit 2 Cycle 24 614.17 Braidwood Unit 1 Cycle 24 620.02 Braidwood Unit 2 Cycle 24 615.05 Table 9-3: Comparison of Fuel Temperature Specification Value K

F Fuel Temperature Used in the Analysis 1050 1430.33 Maximum Average Fuel Temperature for GAIA Transition 1186 Average Fuel Temperature Used in Analysis of Record 977.6 Based on NEI 12-16 (Ref. [9]), treatment of the soluble boron as a burnup averaged value results in the same effect on the fuel reactivity as modeling the actual boron concentration changes as a function of time for complete cycles. Therefore, a conservative soluble boron concentration higher than the maximum burnup-weighted cycle-averaged from the input data is confirmed and used in the depletion calculations, as shown in Table 9-4. Although the references of soluble boron concentrations shown in Table 9-4 are primarily for Braidwood Unit 1, the similarities between the four Braidwood and Byron units allow for the specifications to be representative of all Braidwood and Byron units. Note that the burnup-weighted cycle averaged soluble boron concentration is calculated using the following formula:

=

(+ )()

2

Where, C - burnup-weighted cycle-averaged boron concentration, ppm.

Bui - burnup at the step i Bui+1 - burnup at the step i+1 Bun+1 - burnup at the final step Ci - soluble boron concentration at burnup Bui Ci+1 - soluble boron concentration at burnup Bui+1 Table 9-4: Comparison of Soluble Boron Concentration

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Page 20 of 36 Specification Value Soluble Boron Concentration (ppm) 1000 Burnup-Weighted Cycle Averaged Boron Concentration (ppm)

Braidwood Unit 1 Cycle 21 709 Braidwood Unit 1 Cycle 22 707 Braidwood Unit 1 Design Cycle N 792 Braidwood Unit 1 Design Cycle N+1 915 Sensitivity studies were also performed using the 2x2 array of base case fuel assembly and Region 2 racks to show the reactivity effect of each core operating parameters and confirm the used values were conservative. Separate depletion calculations were performed for different core operating parameters. The results are shown in Table 9-5.

Table 9-5: Results for Core Operating Parameter Cases Description keff Uncertainty Delta-k 2.0 wt%, 5 GWd/MTU Base Case 0.92306 0.00030 Decrease Fuel Temperature by 200K 0.92149 0.00033

-0.00157 Increase Fuel Temperature by 200K 0.92340 0.00032 0.00034 Decrease Moderator Temperature by 100K 0.91903 0.00033

-0.00403 Increase Moderator Temperature by 100K 0.92361 0.00032 0.00055 Decrease Soluble Boron by 200 ppm 0.92171 0.00031

-0.00135 Increase Soluble Boron by 200 ppm 0.92414 0.00031 0.00108 Decrease Specific Power by 5 MW/MTU 0.92277 0.00035

-0.00029 Increase Specific Power by 5 MW/MTU 0.92243 0.00032

-0.00063 3.5 wt%, 20 GWd/MTU Base Case 0.96825 0.00032 Decrease Fuel Temperature by 200K 0.96672 0.00034

-0.00153 Increase Fuel Temperature by 200K 0.96931 0.00035 0.00106 Decrease Moderator Temperature by 100K 0.96183 0.00036

-0.00642 Increase Moderator Temperature by 100K 0.97011 0.00035 0.00186 Decrease Soluble Boron by 200 ppm 0.96757 0.00034

-0.00068 Increase Soluble Boron by 200 ppm 0.96934 0.00038 0.00109 Decrease Specific Power by 5 MW/MTU 0.96910 0.00035 0.00085 Increase Specific Power by 5 MW/MTU 0.96824 0.00036

-0.00001 5.0 wt%, 35 GWd/MTU Base Case 0.98023 0.00034 Decrease Fuel Temperature by 200K 0.97751 0.00035

-0.00272 Increase Fuel Temperature by 200K 0.98200 0.00034 0.00177 Decrease Moderator Temperature by 100K 0.97107 0.00036

-0.00916 Increase Moderator Temperature by 100K 0.98295 0.00037 0.00272 Decrease Soluble Boron by 200 ppm 0.97886 0.00034

-0.00137 Increase Soluble Boron by 200 ppm 0.98134 0.00036 0.00111

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Page 21 of 36 Description keff Uncertainty Delta-k Decrease Specific Power by 5 MW/MTU 0.97968 0.00032

-0.00055 Increase Specific Power by 5 MW/MTU 0.97980 0.00035

-0.00043 Question 10 With respect to integral and fixed burnable absorbers, the licensee states in its LAR, Gadolinia burnable absorbers were also conservatively neglected. The LAR does not clearly specify what this means. NEI 12-16 R4 provides guidance on conservative modeling of gadolinium as an integral burnable absorber, but the LAR does not indicate whether the guidance in NEI 12-16 R4 was followed or if some other modeling approach was taken. For the NRC staff to complete its review, it needs to understand the approach the licensee took with respect to modeling gadolinium as an integral burnable absorber and the impacts of the calculation on keff.

Therefore, the NRC staff requests the following information:

Information related to the details regarding how Gadolinia burnable absorbers were also conservatively neglected.

Justification/applicability of this approach to the Byron and Braidwood Units, Whether or not future operation with the fuel changes affect the conservatism and/or justification/applicability, including:

Change to GAIA AFM fuel, transition to 24-month cycles, and U235 enrichment up to 6.5 wt/%.

Constellation Response to Question 10 Assembly inserts (WABAs, BPRAs, and/or Pyrex) are not planned to be used in any GAIA fuel assembly during depletion; the only credible burnable absorber for GAIA fuel is UO2-Gd2O3 (gadolinia). Since gadolinia is integral to the fuel rods, while spectral hardening does occur, the positive reactivity impact of this effect is never larger than the negative reactivity impact due to residual gadolinium and displacement of fissile material (UO2).

Consequently, it is conservative to neglect the presence of gadolinia in the model and use the maximum fuel density from the fuel without any integral absorbers. Previous studies performed by Oak Ridge National Laboratory (Refs. [4] and [5]) have also shown that neglecting gadolinia burnup absorbers in the criticality analysis is conservative, and the negative reactivity effect yielded by gadolinia bearing rods increases with increasing gadolinia loading and increasing initial fuel enrichment.

As CEG is no longer pursuing greater than 5.0 wt% U-235 fuel, neither GAIA AFM fuel, 24-month cycles, nor enrichments up to 6.5 wt% need to be considered.

Question 11 In its LAR, Constellation did not identify to what extent it operates with control rods inserted. NEI 12-16 R4 states, The criticality safety analysis should include the impact of exposure to fully or partially inserted control rods (and/or part length rods) since rodded operation typically

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Page 22 of 36 increases the fuel assembly reactivity at a given burnup For the NRC staff to complete its review it needs to understand the approach the licensee used with respect to modeling its rodded operation and how it impacts the calculation of keff. Therefore, the NRC staff requests information regarding what operation, if any, the licensee expects the plant to experience with partially or fully inserted control rods, and how this is addressed in the criticality safety analyses.

Constellation Response to Question 11 Control rods comprised of Silver-Indium-Cadmium (Ag-In-Cd) inside stainless steel cladding are planned to be used for short term reactivity control in the reactor core at Braidwood and Byron.

It is expected that control rods will be completely outside of the active fuel region for the entire cycle, except that Bank D used a BEACON bite position of 220 steps at Braidwood and 221 steps at Byron which may position the control rods below the top of the active fuel. Based on the reload schedule and energy requirement, the analytic all rods out (ARO) position is 224 steps and long term control rod insertion position is not below 215 steps (insertion depth is about 6 inches).

Therefore, a sensitivity study was performed with an assumption of an insertion of 8 inches for Ag-In-Cd control rods, corresponding to the top axial segment in the axial burnup distribution, which is more conservative than the maximum insertion depth at the full power condition. The maximum number (24) of control rods was also assumed to be inserted in the top axial segment for the entire fuel irradiation history. The reactivity difference between the keff of the cases with the control rods insertion and the base case keff was considered as the bias, and the 95/95 uncertainty of the bias was statistically combined with other uncertainties. The results are shown in Table 11-1.

Table 11-1: Results for Control Rods Insertion Cases Description keff Uncertainty Delta-k 95/95 unc 2.0 wt%, 5 GWd/MTU Base Case 0.92306 0.00030 Control Rods Insertion 0.92296 0.00031

-0.00010 0.00086 3.5 wt%, 20 GWd/MTU Base Case 0.96825 0.00032 Control Rods Insertion 0.97109 0.00036 0.00284 0.00096 5.0 wt%, 35 GWd/MTU Base Case 0.98023 0.00034 Control Rods Insertion 0.98335 0.00035 0.00312 0.00098 Question 12 In its LAR, the licensee states, 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 [pressurized-water reactor] Special Nuclear Material and Core Component Move Sheet Development. However, the licensee did not specify the meaning of the phrase...isolated from any other fuel assembly. Nor did the licensee describe the methodology that would be used to evaluate reconstituted/non-standard fuel assemblies.

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Page 23 of 36 The LAR did not include information related to the justification for the treatment of reconstituted/non-standard fuel assemblies.

Therefore, the NRC staff requests the following information:

Provide a detailed analysis of fuel assemblies undergoing reconstitution, Identify any limits on reconstituted/non-standard fuel assemblies, Propose appropriate licensing controls to ensure those limits are adhered to.

Constellation Response to Question 12 Currently, there are no GAIA fuel assemblies that need fuel reconstitution at Braidwood or Byron. Therefore, no evaluation has been performed for a fuel assembly with reconstitution in this analysis.

In the case that a fuel assembly is identified as requiring reconstitution in the future, it will be repaired in the fuel elevator with the assembly isolated from any other fuel assemblies.

Sufficient space around the reconstitution location will be ensured to preclude criticality concerns. In addition, Boral panels are attached on the exterior cell walls in the rack cut-out areas around the fuel elevator to preclude a criticality concern.

During the reconstitution activities, missing fuel rods in the fuel assembly may be replaced with dummy rods. For any reconstituted fuel assembly with replacement dummy rods and no empty fuel rod locations, the negative reactivity effect due to reduced amount of fissile material is dominant. Thus, it is always bounded by the base case assembly analyzed and no further evaluation is required.

The fuel rod may be also removed without being replaced by a dummy rod during fuel reconstitution. In this case, especially when there are missing fuel rods in the inner regions of fuel assembly, the reactivity effect from the increase of moderator may be larger than the effect from the loss of fissile material, thus the reactivity of an undermoderated fuel assembly may increase. A previous study performed by Oak Ridge National Lab (Ref. [8]) shows that the reconstituted fuel assembly with loss of multiple fuel rods up to 10% in total in the inner regions of the assemblies would have the maximum reactivity increase up to 0.015 with guide/water hole tubes present, and up to 0.019 with the guide/water hole tubes replaced with water.

However, the actual number of rods removed from a fuel assembly is expected to be much less than the most reactive configuration discussed in Ref. [8]. In addition, the fuel movement in the fuel elevator, including reconstitution activity, is well bounded by the case of a single fresh fuel assembly in water, for which fresh fuel is considered instead of burned fuel that needs to be reconstituted, and no Boral panel around the fuel elevator was credited conservatively.

Calculations were performed for a single isolated fresh fuel assembly at 5.0 wt% without any gadolinia rods in unborated water, and results in Table 12-1 show that there is still plenty of margin to the regulatory limit to offset a reactivity increase due to missing rods from reconstitution.

Table 12-1: Results of Single Isolated Fuel Assembly Cases Description Enrichment (wt%)

Burnup (GWd/MTU) keff Uncertainty Kmax(95/95)

(Note 1)

Single Assembly in Fresh Water 5.0 0.00 0.93769 0.00040 0.95203

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Page 24 of 36 Note 1: The applicable biases and uncertainties determined for fresh fuel from Table 8.2.15 of

[2] is used to determine kmax (95/95).

No additional restrictions will be necessary for reconstituted fuel assembly with replaced dummy rods to be stored in the spent fuel racks, since it is always bounded by the base case assembly.

For the fuel assembly with missing fuel rods to be stored in the spent fuel racks, additional calculations will be performed for each fuel assembly using the same base case model of 2x2 array cells of Ref. [2], but any missing fuel rod will be conservatively replaced with water along the active fuel length. If there is any positive reactivity effect, the fuel assembly may be stored in the Region 1 racks, as the large available safety margin from Region 1 racks may be used to offset the reactivity increase. Otherwise, empty storage cells may be also created in the adjacent locations of the reconstituted assembly for storage in Region 1 or Region 2 racks.

CEG is tracking a revision to procedure NF-AP-309 to make sure all requirements for the fuel movement and execution of reconstitution activities are addressed.

Question 13 In its LAR, page 37, the licensee credits administrative control and process check to preclude analyzing misloading multiple fresh assemblies into Region 2. However, the LAR does not state what administrative control and process check are being relied upon. In this regard, the NRC staff considers the licensees statement to be too vague to be evaluated. The LAR did not Include information related to the administrative controls and process checks being relied upon to preclude the multiple misloading of fresh fuel into Region 2.

Therefore, the NRC staff requests the following information:

The licensees event tree analysis that shows the probability of misloading multiple fresh fuel assemblies into Region 2 or other justification that such an event is not credible.

Constellation Response to Question 13 It is important to have a defense-in-depth program in place to minimize the severity of a scenario where multiple fresh assemblies are loaded into the wrong storage locations. As the first barrier to preclude the multiple misloading of fresh fuel in Region 2, Constellation has procedures which provide guidance for fuel assembly storage in accordance with the Technical Specification and the spent fuel pool criticality analysis, as listed below:

x Constellation Procedure NF-AA-330, "SPECIAL NUCLEAR MATERIAL PHYSICAL INVENTORIES," defines the method to be used at Constellation Nuclear Operated Sites for performing Special Nuclear Material (SNM) inventories to meet NRC Regulations 10CFR74.19(c) and 10CFR72.72(b) x Constellation Procedure NF-AP-309, "PWR Special Nuclear Material and Core Component Move Sheet Development," provides requirements for the control of the movement of Special Nuclear Material (SNM) and Non-Fuel Components at the PWR reactor sites operated by Constellation

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Page 25 of 36 x

Constellation Procedure NF-BR-310-2000, "Special Nuclear Material and Core Component Movement Requirements for Braidwood," provides requirements for the control of the movement of nuclear fuel and other Special Nuclear Material at Braidwood Station x

Constellation Procedure NF-BY-310-2000, "Special Nuclear Material and Core Component Movement Requirements for Braidwood," provides requirements for the control of the movement of nuclear fuel and other Special Nuclear Material at Byron Station x

Constellation Procedure OU-AP-4001, "PWR FUEL AND CORE COMPONENT HANDLING PRACTICES," sets forth expectations for handling nuclear fuel and non-SNM core components at PWR stations using Move Sheets to prevent Station and Outage Services Fuel Handling activities issues Specifically,

1. of NF-AP-309 specifies the burnup requirements for Byron and Braidwood Region 2 racks. The burnup of each assembly needs to be verified before the assembly is placed into Region 2. Section 4.B of Attachment 8 of NF-BR-310-2000 lists detailed requirements to ensure assemblies initially entering the Region 2 cells meet the minimum burnup requirements. Adherence to the requirements in the above procedures is accomplished prior to any fuel movement, therefore the fuel assembly is ensured to be placed in an acceptable location.

2.

To ensure that the correct fuel assembly is selected for placement in an acceptable location, the fuel assembly serial number and pickup and set down locations are verified by the fuel handling operators following Constellation Procedure OU-AP-4001 Attachment 6 of OU-AP-4001 provides the standard script template for PWR spent fuel pool movement that the fuel handling crew follows.

3.

After fuel movement, as discussed in Section 4.7.2 of Constellation Procedure NF-AA-330, a location inventory and a serial number inventory of fuel assemblies residing in the spent fuel pool will be conducted, which provides an independent confirmation of the acceptable storage configurations in the Region 2 racks.

The second barrier to prevent a common-fault error of multiple misloading of fresh fuel assemblies is a visual verification to ensure that a fresh fuel assembly is not selected when a used fuel assembly is intended to be selected based on the physical appearance of the assembly. A description of the physical appearance difference between fresh and burnt fuel is incorporated into Fuel Handler training materials.

Question 14 The licensee performed calculations for the loss of SFP cooling, mislocated fuel assembly, single misloaded fuel assembly, and the multiple misload accidents. In its LAR, Table 10, the licensee provided bounding results (Kmax) from the calculations for mislocated fuel assembly, single misloaded fuel assembly, and the multiple misload accidents. Bounding results (Kmax) were not provided for the loss of SFP cooling.

The NRC staff requests the details of the temperature sensitivity study performed to support the licensee's statement in the LAR that the 4 degrees Celsius moderator temperature is bounding.

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Page 26 of 36 Constellation Response to Question 14 Under accident conditions (loss of cooling) the temperature could be elevated beyond the normal operating range for Region 1 or Region 2. TS 3.7.15 specifies lowest SFP water normal operating temperature as 50 °F (10 °C). Braidwood-Byron UFSAR Chapter 9 specifies 145.2 °F (62.9 °C) as the SFP water design temperature, and 165.3 °F (74.1 °C) as the SFP water design temperature (with one cooling train in operation).

By changing the water density (as shown in Table 14-1) and its associated TMP card 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. Furthermore, two sets of calculations were performed for each water temperature, one at the lower bound of thermal WUHDWPHQW6DQGWKHRWKHURQHDWWKHXSSHUERXQGRIWKHUPDOWUHDWPHQW if both thermal treatments were available. Otherwise, only the nearest available thermal treatment was used.

As an example, for water temperature at 60 °C (333.15 K), two MCNP models were run, one lwtr.20t (corresponds to 293.6 K) and the other one lwtr.21t (corresponds to 350 K). The 4 °C moderator temperature was used for the base case model (base case fuel and rack).

The difference between the maximum keff of the moderator-temperature cases and the base case keff was determined. The results for Region 1 and Region 2 are shown in Tables 14-2 and 14-3, respectively. It is shown that the bounding condition is the base case with the minimum temperature (4 °C) and maximum density for both regions.

Table 14-1: Water Density as a Function of Temperature at One Atmospheric Pressure Water Temperature (oC)

Water Density (g/cm3) 4 1

20 0.99823 40 0.99225 60 0.98323 80 0.97182 100 0.95838 124 and 0% void 0.95838 124 and 10% void 0.86254 124 and 20% void 0.76670 Note 1. The water density for 100 °C is assumed for water at 124 °C.

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 27 of 36 Table 14-2: Results for Moderator Temperature and Density Cases, Region 1 Water Temperature and Void Amount LWTR (Note 1) keff Uncertainty Borated Water 4 °C (base case) 20t 0.87206 0.00018 20 °C 20t 0.87160 0.00018 40 °C 21t 0.86922 0.00018 60 °C 21t 0.86701 0.00018 80 °C 22t 0.86303 0.00018 100 °C 22t 0.85933 0.00018 124 °C, 0% void 22t 0.85951 0.00018 124 °C, 10% void 22t 0.83470 0.00018 124 °C, 20% void 22t 0.80853 0.00018 40 °C 20t 0.87003 0.00018 60 °C 20t 0.86754 0.00018 80 °C 21t 0.86387 0.00018 100 °C 21t 0.86087 0.00018 124 °C, 0% void 21t 0.86092 0.00018 124 °C, 10% void 21t 0.83604 0.00018 124 °C, 20% void 21t 0.80953 0.00018 Bias 0.00000 Unborated Water 4 °C (base case) 20t 0.92318 0.00018 20 °C 20t 0.92229 0.00018 40 °C 21t 0.91903 0.00018 60 °C 21t 0.91578 0.00018 80 °C 22t 0.91051 0.00018 100 °C 22t 0.90637 0.00018 124 °C, 0% void 22t 0.90601 0.00018 124 °C, 10% void 22t 0.87529 0.00018 124 °C, 20% void 22t 0.84310 0.00018 40 °C 20t 0.92082 0.00018 60 °C 20t 0.91803 0.00018 80 °C 21t 0.91244 0.00018 100 °C 21t 0.90809 0.00018 124 °C, 0% void 21t 0.90812 0.00018 124 °C, 10% void 21t 0.87714 0.00018 124 °C, 20% void 21t 0.84475 0.00018 Bias 0.00000 Note 1. 20t, 21t and 22t are the ENDF/B-9,,WKHUPDO6FURVVVHFWLRQOLEUDU\\DW

293.6 K (20.45 oC), 350 K (76.85 oC), and 400 K (126.85 oC), respectively.

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 28 of 36 Table 14-3: Results for Moderator Temperature and Density Cases, Region 2 Water Temperature and Void Amount LWTR (Note 1) keff Uncertainty Delta-k 2.0 wt%, 0 GWd/MTU 4 °C 20t 0.96558 0.00032 20 °C 20t 0.96549 0.00033

-0.00009 60 °C 21t 0.95374 0.00031

-0.01184 60 °C 20t 0.96179 0.00032

-0.00379 100 °C 22t 0.94076 0.00032

-0.02482 100 °C 21t 0.94864 0.00031

-0.01694 124 °C 22t 0.94155 0.00031

-0.02403 124 °C 21t 0.94840 0.00031

-0.01718 124 °C, 10% void 22t 0.91934 0.00034

-0.04624 124 °C, 10% void 21t 0.92713 0.00034

-0.03845 124 °C, 20% void 22t 0.89362 0.00030

-0.07196 124 °C, 20% void 21t 0.90027 0.00033

-0.06531 3.5 wt%, 20 GWd/MTU 4 °C 20t 0.96825 0.00032 20 °C 20t 0.96819 0.00035

-0.00006 60 °C 21t 0.96075 0.00036

-0.00750 60 °C 20t 0.96430 0.00034

-0.00395 100 °C 22t 0.95108 0.00034

-0.01717 100 °C 21t 0.95452 0.00035

-0.01373 124 °C 22t 0.95129 0.00035

-0.01696 124 °C 21t 0.95447 0.00035

-0.01378 124 °C, 10% void 22t 0.92634 0.00034

-0.04191 124 °C, 10% void 21t 0.92860 0.00034

-0.03965 124 °C, 20% void 22t 0.89605 0.00036

-0.07220 124 °C, 20% void 21t 0.89833 0.00033

-0.06992 Note 1. 20t, 21t and 22t are the ENDF/B-9,,WKHUPDO6FURVVVHFWLRQOLEUDU\\DW.

(20.45 °C), 350 K (76.85 °C), and 400 K (126.85 °C), respectively.

I I

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 29 of 36 Table 14-3 (Continued): Results for Moderator Temperature and Density Cases, Region 2 Water Temperature and Void Amount LWTR (Note 1) keff Uncertainty Delta-k 5.0 wt%, 35 GWd/MTU 4 °C 20t 0.98023 0.00034 20 °C 20t 0.97944 0.00037

-0.00079 60 °C 21t 0.97297 0.00036

-0.00726 60 °C 20t 0.97475 0.00038

-0.00548 100 °C 22t 0.96450 0.00033

-0.01573 100 °C 21t 0.96612 0.00035

-0.01411 124 °C 22t 0.96472 0.00035

-0.01551 124 °C 21t 0.96660 0.00035

-0.01363 124 °C, 10% void 22t 0.93806 0.00035

-0.04217 124 °C, 10% void 21t 0.93953 0.00037

-0.04070 124 °C, 20% void 22t 0.90637 0.00035

-0.07386 124 °C, 20% void 21t 0.90771 0.00036

-0.07252 Note 1. 20t, 21t and 22t are the ENDF/B-9,,WKHUPDO6) cross section library at 293.6 K (20.45 °C), 350 K (76.85 °C), and 400 K (126.85 °C), respectively.

Question 15 The licensee's description of the NFV in its LAR lacks specificity and justification. The LAR does not identify the NFV dimensions that were used in the model and how they relate to the actual dimensions. The licensee stated in its LAR that it used dimensions less than nominal but did not provide justification or identify the impact on the results. The licensee did not justify or demonstrate the acceptability of not following the NEI 12-16 R4 and RG 1.240 R0 guidance with respect to modeling steel below the active fuel and the NFV concrete walls.

Compliance with 10 CFR 50.68(b)(2) requires that keff of the NFV not exceed 0.95, at a 95 percent probability, 95 percent confidence level assuming the NFV is flooded with full density water. The licensee did not provide sufficient justification for the NRC staff to make an independent evaluation.

Compliance with 10 CFR 50.68(b)(3) requires that keff of the NFV not exceed 0.98, at a 95 percent probability, 95 percent confidence level assuming the NFV is flooded with an optimum density moderator. The licensee did not provide sufficient justification for the NRC staff to make an independent evaluation.

The above information is important to the NRCs review of the licensees estimation of the NFV keff during abnormal/accident conditions.

Therefore, the NRC staff requests the following information:

NFV dimensions that were used in the model and how they relate to the actual dimensions.

I I

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 30 of 36 The effect of modeling the steel below the active fuel and the NFV concrete walls in a manner consistent with NEI 12-16 R4 and Regulatory Guide 1.240 or demonstrate the acceptability of not performing this modeling.

Information related to the licensees NFV fully flooded analysis.

Information related to the licensees NFV optimum moderation analysis.

Constellation Response to Question 15 Modeled NFV cavity dimensions:

The NFV cavity dimensions of 558" x 279" x 173" were modeled, while the actual dimensions are 576" x 284" x 173". Other NFV dimensions are included in Table 15-1.

Table 15-1: NFV Major Specifications Specification Value Box ID (in.)

9 + 0.12 / - 0 Box wall thickness (in.)

0.125 +/- 0.01 Box height (in.)

168.25 Cell pitch (in.)

21 +/- 0.06 Distance between two arrays (in.)

50.87 +/- 0.06 NFV basis thickness (in.)

7 NFV height (in.)

(Note 1) 180 Note 1. The NFV height of 180 inches in this table includes 7 inches NFV basis thickness.

A supplementary calculation was performed to evaluate the effect of the difference between the modeled NFV cavity dimensions to the actual NFV cavity dimensions. The result is provided in Table 15-2.

Table 15-2: Reactivity Effect of Modeled NFV with Actual Dimensions Description keff Uncertainty Full-moderator density Base case 0.89916 0.00018 NFV with actual cavity dimensions 0.89895 0.00018 Bias 0.00000 Optimum-moderation water density Base case 0.72691 0.00016 NFV with actual cavity dimensions 0.72505 0.00015 Bias 0.00000 As can be seen, the deviation does not affect keff of the full-moderator density case, but results in a lower keff of the optimum-moderation water density case.

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 31 of 36 Concrete composition:

The concrete composition for design base models was from SCALE 6.2 manual, Oak Ridge concrete. This concrete composition was selected since it has a low hydrogen (H) content, which results in a relatively large reactivity effect compared to compositions with a higher hydrogen amount. Nevertheless, an additional sensitivity study was performed and is discussed below.

Three concrete compositions and densities were evaluated, with no other changes to the design base model, as:

x The dry conservative concrete composition and density from Table 10-5 of the EPRI "Sensitivity Analysis for Spent Fuel Pool Criticality - Revision 1" (Ref. [6]), with no hydrogen x

The wet conservative concrete composition and density from Table 10-5 of the EPRI sensitivity analysis (Ref. [6]), with a hydrogen amount of 0.26 wt%

x Average of the above concrete compositions, with a hydrogen amount of 0.13 wt%

The results of this study are provided in Table 15-3 and show that the concrete composition does not change the reactivity for the full-moderator density, but the reactivity effect of EPRI dry concrete composition is bounding for optimum-moderator density. The EPRI dry concrete composition was used in the following study and the result was used as a bias to determine kmax(95/95).

Table 15-3: Reactivity Effect of using EPRI Concrete for NFV Walls Description keff Uncertainty Full-moderator density Base case 0.89916 0.00018 EPRI dry conservative concrete 0.89916 0.00018 EPRI wet conservative concrete 0.89916 0.00018 Average of wet and dry concrete compositions 0.89916 0.00018 Bias 0.00000 Optimum-moderation water density Base case 0.72691 0.00015 EPRI dry conservative concrete 0.73708 0.00015 EPRI wet conservative concrete 0.73201 0.00015 Average of wet and dry concrete compositions 0.73410 0.00015 Bias 0.01017

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 32 of 36 EPRI Dry Concrete Composition, NFV Base and Perimeter Modeling, and Moderator Below and Above Fuel Active Region:

Another sensitivity analysis is performed with the EPRI dry conservative concrete, but with the following changes:

x 36" concrete is modeled as perimeter of the NFV, and as the NFV base material.

x The actual NFV cavity is modeled.

x The rack box below and above fuel active region replaced with the moderator.

The result of this study is shown in Table 15-4. The positive difference between this sensitivity study and the design basis was used as the "geometry and material modeling bias" when calculating kmax(95/95) as shown in Table 1 of Attachment 1.

Table 15-4: Reactivity Effect of NFV with Actual Cavity Dimensions, and with EPRI Dry Conservative Concrete for NFV Walls and Floor Description keff Uncertainty Full-moderator density Base case 0.89916 0.00018 NFV with actual cavity dimensions, with EPRI dry concrete for NFV 36" walls and floor 0.89927 0.00018 Bias 0.00011 Optimum-moderation water density Base case 0.72691 0.00015 NFV with actual cavity dimensions, with EPRI dry concrete for NFV 36" walls and floor 0.76579 0.00015 Bias 0.03888 NFV fully flooded analysis:

The nominal values for GAIA 17x17 fuel were used to build the MCNP model of the base case fuel assembly, except the active region was modeled with the fuel enrichment of 5.0 wt%.

Except for length and width of the NFV cavity, the nominal values were used to build the MCNP model of the base case NFV racks. As discussed above (under Modeled NFV cavity dimensions), a sensitivity analysis performed that showed the length and width of the NFV cavity were modeled conservatively. Another sensitivity study was performed as discussed above (under EPRI Dry Concrete Composition, NFV Base and Perimeter Modeling, and Moderator Below and Above Fuel Active Region) by replacing the rack box below and above fuel active region with the moderator, the difference was added as a bias as "Geometry and material modeling bias" in Table 1.

The following shows the list of biases and uncertainties considered to calculate kmax(95/95).

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 33 of 36 Biases:

x Criticality code validation bias x

Eccentric positioning bias x

Geometry and material modeling bias x

Temperature reactivity effect Uncertainties:

The applicable uncertainties are provided below. This list covers applicable uncertainties provided in NEI 12-16 (Ref. [9]).

x Fuel manufacturing tolerances (95/95)

R Fuel pitch R

Fuel pellet diameter R

Fuel clad ID R

Fuel clad OD R

Guide tube ID R

Guide tube OD R

Fuel enrichment R

Fuel density x

Rack manufacturing tolerances (95/95)

R Box ID R

Box wall thickness R

Cell pitch R

Distance between two arrays x

Criticality code validation uncertainty 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.

The reactivity effect of some other parameters, such as temperature, is also evaluated. If the reactivity effect of a parameter was positive, the keff difference was added as a bias when calculating the kmax(95/95).

NFV optimum moderation analysis:

The base case model for optimum moderation analysis is the same as the model for the fully flooded analysis, except for the moderator density. 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.

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 34 of 36 The bounding moderator density for optimum moderation analysis was 0.04 g/cm3 and used in further analysis. The evaluated biases and uncertainties were the same as those discussed above in the NFV fully flooded analysis discussion.

A major difference between optimum moderation and fully flooded result is the temperature reactivity effect. While for the fully flooded NFV, the base case with 4 °C is bounding, for the optimum moderator density, the maximum reactivity is at 124 °C. At this temperature, three other water densities of 0.03, 0.06 and 0.15 g/cm3 were evaluated which results are shown in Table 15-5. It is shown that the maximum reactivity is at 124 °C and water density of 0.04 g/cm3.

The difference between this bounding case and the base case is positive, thus it was added as a bias when calculating kmax(95/95), as shown in Table 1 of Attachment 1.

Table 15-5: Temperature Reactivity Effect T

(oC)

Water Density (g/cm3)

LWTR keff Uncertainty Full-moderator density 4 (Base case) 1 20t 0.89916 0.00018 20 0.99823 20t 0.89841 0.00018 80 0.97182 21t 0.89491 0.00018 124 0.95838 22t 0.89453 0.00018 Bias 0.00000 Optimum-moderation water density 4 (Base Case) 0.04 20t 0.72691 0.00015 20 0.04 20t 0.72686 0.00015 80 0.04 21t 0.74397 0.00015 124 0.04 22t 0.75659 0.00015 124 0.06 22t 0.75657 0.00015 124 0.15 22t 0.72049 0.00015 124 0.03 22t 0.73834 0.00015 Bias 0.02968 Question 16 Pursuant to 10 CFR 50.36(c)(4) it requires design features of the facility such as materials of construction and geometric arrangements to be included in TS, which, if altered or modified, would have a significant effect on safety. The Byron and Braidwood TSs currently does not have any design features TS for its NFV. The licensee has proposed a change to its NFV Analysis of Record (AoR) that would accommodate a U-235 enrichment of 6.5 wt/%. This is a 30 percent increase in fissile material over its current approved AoR. Given this significant change, the NRC staff requests the licensee demonstrate that it continues to meet 10 CFR 36(c)(4).

Constellation Response to Question 16 10 CFR 50.36 implements Section 182a of the Atomic Energy Act (AEA) of 1954 (Public Law 83--703) that requires the inclusion of Technical Specifications (TS) in the operating license of a nuclear power plant. Section 182a, "License Applications," states, in part, that the TS shall

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 35 of 36 include "information of the amount, kind, and source of special nuclear materials required, the place of use, [and] the specific characteristics of the facility". The TS Chapter 4.0, "Design Features," discusses those features of a nuclear power plant, such as materials of construction and geometric arrangement, which if altered or modified, could have a significant impact on safety. Historically, TS Chapter 4.0 included discussion of the following topics:

x The site location (4.1);

x Fuel assemblies (4.2.1);

x Control rods (4.2.2);

x Spent fuel storage criticality protection (4.3.1);

x New fuel storage criticality protection (typically in 4.3.1, if included);

x Spent fuel pool drainage protection (4.3.2); and x

Spent fuel pool capacity (4.3.3).

As part of the conversion from custom TS to ITS, some details previously included in the Design Features chapter were relocated to the UFSAR. Both the Braidwood and Byron Chapter 4 TS contained the above sections at the time their TS were converted to the ITS, issued as Amendments 98 and 106 (ADAMS Accession No. 9901040081) on December 22, 1998, respectively, except that the new fuel vault criticality protection was not included Section 4.3 at the time these amendments were issued and has not been added since that time.

Since the Braidwood and Byron TS do not currently contain any information about new fuel vault (NFV) criticality protection and CEG is no longer requesting approval for storing greater than 5.0 wt% U-235 fuel in the NFV; there is no need to add a new NFV requirement to TS 4.0 as part of this proposed amendment request.

ATTACHMENT 2 Response to Request for Additional Information (Non-Proprietary)

Page 36 of 36 References

[1]

BRW-24-0031-N/BYR24-011, Revision 0, "Braidwood-Byron GAIA Criticality Analysis for New Fuel Vault", dated October 2024.

[2]

BRW-24-0032-N/BYR24-012 Revision 0, "Braidwood-Byron Criticality Analysis for Spent Fuel Pool", dated October 2024.

[3]

NUREG/CR-6800, ORNL/TM-2002/6, Assessment of Reactivity Margin and Loading Curves for PWR Burnup-Credit Cask Designs, dated March 2003. (ADAMS Accession No. ML031110280)

[4]

ORNL/TM-12973, Sensitivity and Parametric Evaluations of Significant Aspects of Burnup Credit for PWR Spent Fuel Packages," dated May 1996. (ADAMS Accession No. ML011440229)

[5]

NUREG/CR-6760, ORNL/TM-2000/321, Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit, dated March 2002. (ADAMS Accession No. ML020770436)

[6]

EPRI Report 3002008197, "Sensitivity Analyses for Spent Fuel Criticality - Revision 1",

dated November 2017 (ADAMS Accession No. ML18088B399).

[7]

NRC Regulatory Guide 1.183, Rev 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", dated July 2000.

(ADAMS Accession No. ML003716792)

[8]

NUREG/CR-6835, ORNL/TM-2002/255, "Effects of Fuel Failure on Criticality Safety and Radiation Dose for Spent Fuel Casks," dated September 2003. (ADAMS Accession No. ML032880058)

[9]

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)

ATTACHMENT 3 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 Braidwood Mark-up of Technical Specifications Page

Spent Fuel Assembly Storage 3.7.16 BRAIDWOOD UNITS 1 & 2 3.7.16 1 Amendment 145 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Assembly Storage LCO 3.7.16 Each spent fuel assembly stored in the spent fuel pool shall, as applicable:

a.

Region 1 of Holtec spent fuel pool storage racks Have an initial nominal enrichment of 5.0 weight percent U-235 to permit storage in any cell location.

b.

Region 2 of Holtec spent fuel pool storage racks Have a combination of initial enrichment and burnup within the Acceptable Burnup Domain of Figure 3.7.16-1.

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.

Requirements of the LCO not met.

A.1 Initiate action to move the noncomplying fuel assembly into a location which restores compliance.

Immediately delete delete I

Spent Fuel Assembly Storage 3.7.16 BRAIDWOOD UNITS 1 & 2 3.7.16 2 Amendment 145 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means the initial nominal enrichment of the fuel assembly is 5.0 weight percent U-235.

Prior to storing the fuel assembly in Region 1 SR 3.7.16.2 Verify by administrative means the combination of initial enrichment, burnup, and decay time, as applicable, of the fuel assembly is within the Acceptable Burnup Domain of Figure 3.7.16-1.

Prior to storing the fuel assembly in Region 2 and delete

Spent Fuel Assembly Storage 3.7.16 BRAIDWOOD UNITS 1 & 2 3.7.16 3 Amendment 145 Figure 3.7.16-1 (page 1 of 1)

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

Figure being replaced by figure on next page delete 5000

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ATTACHMENT 4 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 Byron Mark-up of Technical Specifications Page

Spent Fuel Assembly Storage 3.7.16 BYRON UNITS 1 & 2 3.7.16 1 Amendment 165 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Assembly Storage LCO 3.7.16 Each spent fuel assembly stored in the spent fuel pool shall, as applicable:

a.

Region 1 of spent fuel pool storage racks Have an initial nominal enrichment of 5.0 weight percent U-235 to permit storage in any cell location.

b.

Region 2 of spent fuel pool storage racks Have a combination of initial enrichment and burnup within the Acceptable Burnup Domain of Figure 3.7.16-1.

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.

Requirements of the LCO not met.

A.1 Initiate action to move the noncomplying fuel assembly into a location which restores compliance.

Immediately 12 &+$1*(6 0$'( 72 7+,6 3$*(

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Prior to storing the fuel assembly in Region 1 SR 3.7.16.2 Verify by administrative means the combination of initial enrichment and burnup, as applicable, of the fuel assembly is within the Acceptable Burnup Domain of Figure 3.7.16-1.

Prior to storing the fuel assembly in Region 2 12 &+$1*(6 0$'( 72 7+,6 3$*(

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Spent Fuel Assembly Storage 3.7.16 BYRON UNITS 1 & 2 3.7.16 3 Amendment 165 INITIAL U-235 ENRICHMENT (w/o)

Figure 3.7.16-1 (page 1 of 1)

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ATTACHMENT 5 Revised NEI 12-16 Checklists

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 1 of 13 Compliance with NEI 12-16, Benchmarking Calculations:

The checklist responses are unchanged from Attachment 6 of the original submittal.

(Attachment 1, Ref. 4.1.1).

Compliance with NEI 12-16, NFV Criticality Calculations:

The section, table and figure numbers referenced in the following table are specific to the NFV Criticality Calculation Design Analysis (Attachment 1, Reference 4.2.1).

Subject Included Notes / Explanation 1.

Introduction and Overview Purpose of submittal Yes Section 1 Changes requested Yes A new analysis is performed introducing a new fuel (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.

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 2 of 13 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 Control 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.

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 3 of 13 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 keff 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, and Attachment C Fuel density Yes Section 8.2 Burnable Poisons No Not Applicable Fuel dimensions Yes Section 8.4 Axial blankets Yes Attachment C, in a sensitivity analysis Limiting rack model Section 8.5 Storage vault dimensions and materials Yes Attachment C Temperature Yes Attachment C

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 4 of 13 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 Integral absorber No Not modeled.

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, and 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

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 5 of 13 9.

Criticality Safety Analysis of Spent Fuel Pool Storage Racks No Not applicable 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

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 6 of 13 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

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 7 of 13

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:

The section, table and figure numbers referenced in the following table are specific to the SFP Criticality Calculation Design Analysis (Attachment 1, Ref. 4.2.2).

Subject Included Notes / Explanation 1.

Introduction and Overview Purpose of submittal Yes Section 1 Changes requested Yes A new analysis is performed introducing a new fuel (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

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 8 of 13 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, Table 3.3.1 Schematic (axial/cross-section)

Yes Figures 7.1.1 and 7.2.1 show radial cross section. No axial variation in modeling.

Material compositions Yes Table 3.2.3 Other Reactivity Control 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.

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 9 of 13 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 keff 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

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 10 of 13 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

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 11 of 13 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

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 12 of 13 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 Yes Section 1 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.

ATTACHMENT 5 Revised NEI 12-16 Checklists Page 13 of 13 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 6 Constellation Affidavit for Withholding Constellation for Attachments 11 and 12

Page 1 of 2 AFFIDAVIT 1.

My name is Rebecca L. Steinman. 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 Attachments 11 and 12 to CEG letter RS-24-098 dated October 23, 2024, with subject "Supplement to License Amendment Request to Revise 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 and is bracketed and marked with a C superscript to indicate CEG proprietary 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

Page 2 of 2 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: (10/23/2024) rebeccca.steinman@constellation.com 4300 Winfield Road Warrenville, IL 60555

Steinman, Rebecca Lee Digitally signed by Steinman, Rebecca Lee Date: 2024.10.23 18:10:07 -05'00'

ATTACHMENT 7 Holtec Affidavit for Withholding for Attachment 11

U.S. Nuclear Regulatory Commission A TIN: Document Control Desk Constellation Letter RS-24-098 Non-Proprietary Attachment 7 AFFIDAVIT PURSUANT TO 10 CFR 2.390 I, Kimberly Manzione, being duly sworn, depose and state as follows:

( 1)

I have reviewed the information described in paragraph (2) which is sought to be withheld, and am authorized to apply for its withholding.

(2)

The information sought to be withheld is the information within brackets, identified with "4.a" or "4.b" superscripts in Attachment 11 to Constellation Letter RS-24-098 (HI-982094), which is Holtec Proprietary information, for the corresponding reasons in item ( 4) below.

(3)

In making this application for withholding of proprietary information of which it is the owner, Holtec International relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4) and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10CFR Part 9.17(a)(4), 2.390(a)(4), and 2.390(b)(l) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4 ). The material for which exemption from disclosure is here sought is all "confidential commercial information",

and some portions also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

I of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Constellation Letter RS-24-098 Non-Proprietary Attachment 7 AFFIDAVIT PURSUANT TO 10 CFR 2.390

( 4)

Some examples of categories of information which fit into the definition of proprietary information are:

a.

Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Holtec's competitors without license from Holtec International constitutes a competitive economic advantage over other companies;

b.

Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c.

Information which reveals cost or price information, production, capacities, budget levels, or commercial strategies of Holtec International, its customers, or its suppliers;

d.

Information which reveals aspects of past, present, or future Holtec International customer-funded development plans and programs of potential commercial value to Holtec International;

e.

Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4.a and 4.b above.

(5)

The information sought to be withheld is being submitted to the NRC in confidence. The information (including that compiled from many sources) is of a sort customarily held in confidence by Holtec International, and is in fact so held. The information sought to be withheld has, to the best of my know ledge and belief, consistent! y been held in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to 2 of5

U.S. Nuclear Regulatory Commission A TIN: Document Control Desk Constellation Letter RS-24-098 Non-Proprietary Attachment 7 AFFIDAVIT PURSUANT TO 10 CFR 2.390 regulatory prov1s10ns or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

( 6)

Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within Holtec International is limited on a "need to know" basis.

(7)

The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function ( or his designee ), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside Holtec International are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8)

The information classified as proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information is classified as proprietary because it contains detailed descriptions of analytical approaches and methodologies not available elsewhere. This information would provide other parties, including competitors, with information from Holtec International's technical database and the results of evaluations performed by Holtec International. A substantial effort has been expended by Holtec International to develop this information. Release of this information would improve a competitor's position because it would enable Holtec' s competitor to copy our technology and off er it for sale in competition with our company, causing us financial lilJUry.

3 of5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Constellation Letter RS-24-098 Non-Proprietary Attachment 7 AFFIDAVIT PURSUANT TO 10 CFR 2.390 (9)

Public disclosure of the information sought to be withheld is likely to cause substantial harm to Holtec International's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Holtec International's comprehensive spent fuel storage technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology, and includes development of the expertise to determine and apply the appropriate evaluation process.

The research, development, engineering, and analytical costs comprise a substantial investment of time and money by Holtec International.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

Holtec International's competitive advantage will be lost if its competitors are able to use the results of the Holtec International experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to Holtec International would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Holtec International of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

4 of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Constellation Letter RS-24-098 Non-Proprietary Attachment 7 AFFIDAVIT PURSUANT TO 10 CFR 2.390 STATE OF NEW JERSEY COUNTY OF CAMDEN

)

)

)

ss:

Kimberly Manzione, being duly sworn, deposes and says:

That she has read the foregoing affidavit and the matters stated therein are true and correct to the best of her know ledge, information, and belief.

Executed at Camden, New Jersey, this 22nd day of October, 2024.

/ /JA.

irnb~rl~ M~

ne Director of Licensing Holtec International One Holtec Boulevard Camden, NJ 08104 Subscribed and sworn before me this,),JJ'~ ay of Q c..-tbk, 2024.

Erika Grandrimo 7

NOTARY PUBLIC STATE OF NEW JERSEY MY COMMISSION EXPIRES JANUARY 17, 2027 5 of 5

ATTACHMENT 8 Westinghouse Affidavit for Withholding for Attachment 11

• • • This record was final approved on 10/23/2024 16:12:53. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-24-053 Page 1 of 3 Commonwealth of Pennsylvania:

County of Butler:

(1)

I, Zachary Harper, Consulting Engineer, Licensing, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2)

I am requesting the proprietary portions of Holtec Report HI-982094 Revision 6 in Constellation Energy Generation Letter RS-24-098 be withheld from public disclosure under 10 CFR 2.390.

(3)

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

(4)

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

(i)

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

(ii)

The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouses knowledge, is not available in public sources.

(iii)

Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

• • • This record was final approved on 10/23/2024 16:12:53. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-24-053 Page 2 of 3 (5)

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

(a)

The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b)

It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c)

Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d)

It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e)

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

(f)

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

(6)

The attached documents are bracketed and marked with a W superscript to indicate the Westinghouse proprietary information. The justification for withholding is identified in Sections (5)(a) and (c) of this Affidavit.

• • • This record was final approved on 10/23/2024 16:12:53. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-24-053 Page 3 of 3 I declare that the averments of fact set forth in this Affidavit 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: 10/23/2024 Signed electronically by Zachary Harper 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066

• • • This record was final approved on 10/23/2024 16:12:53. (This statement was added by the PRIME system upon its validation)

CAW-24-053 Revision 0 Non-Proprietary Class 3

• • This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**

Approval Information Manager Approval Harper Zachary S Oct-23-2024 16:12:53 Files approved on Oct-23-2024

ATTACHMENT 9 Summary of Regulatory Commitments

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 NRC's information and are not regulatory commitments.)

COMMITMENT COMMITTED DATE OR "OUTAGE" COMMITMENT TYPE ONE-TIME ACTION (Yes/No)

PROGRAMMATIC ACTION (Yes/No)

CEG (Braidwood) will relocate the following fuel assemblies into the Region 1 racks. Assembly IDs:

D82U, D73U, D81U, and D77U To be implemented prior to implementation of the license amendment Yes No CEG (Byron) will relocate fuel assembly T77K to Region 1 racks.

To be implemented prior to approval and implementation of the GAIA criticality analysis and updated Byron TS 3.7.16 Yes No

ATTACHMENT 10 HI-982094, Revision 6, "Criticality Evaluation for the Byron/Braidwood Rack Installation Project", dated October 2024 (Non-Proprietary)

Exelon Design Analysis HI-982094 Rev. 5 Page 6 of 111 Braidwood criticality design margin evaluation and disposition - HI-982094 Rev. 5 Spent fuel pool criticality margin was evaluated for impact based on Information Notice (IN) 2011-03 identified issues under OPEX Review assignment #1206369-01. The IN discussed two issues, of which only one was applicable to Braidwood. The IN stated that the standard deviation of the population about the mean is the correct statistical method for determining the Monte Carlo code bias uncertainty documented in NUREG/CR-6698. Braidwoods criticality analysis of record (AOR) HI-982094 used uncertainty of the mean. The impact is explicitly calculated below (the source of the calculational spreadsheet is documented as an attachment to OPEX Review assignment #1206369-01):

Table 1 Criticality Analysis Maximum keff Rack Up - IN 2011-3 Penalty Byron/Braidwood (HI-982094)

Current AOR Corrected Region 1

Region 2

Region 1

Region 2

Reference keff 0.9318 0.9129 0.9318 0.9129 Uncertainties Bias Uncertainty

[

]4a,4b Calculational Statistics

[

]C Depletion Uncertainty

[

]C Fuel Eccentricity 0.0000 0.0000 0.0000 0.0000 Manufacturing Tolerances

[

]C Statistical Combination of Uncertainties

[

]C Biases Axial Burnup Distribution

[

]C Calculational Bias

[

]C Maximum Base keff 0.9422 0.9377 0.9451 0.9396 Penalties Mark BW LUA 0.0000 0.0071 0.0000 0.0071 Maximum Total keff 0.9422 0.9448 0.9451 0.9467 keff increase due to population vs.

mean 0.0029 0.0019 The results above show that the impact is small (~3mk). Therefore, with the impact included in the calculation, there is still sufficient margin to the limit of 0.95. The corrected keff, accounting for the IN 2011-3 penalty, is 0.9451 for Braidwood Region 1 and 0.9467 for Braidwood Region 2.

Report

Title:

Criticality Evaluation for the Byron/Braidwood Rack Installation Project Holtec Report # 982094 Report Type: Project Specific Holtec Project # 80944 Name of Contract Holder Holtec Enterprise Unit: Nuclear Power Division Client Name: Constellation Energy Nuclear Safety Category: Safety Significant Confidentiality Class1: Non-Proprietary Information on Names of Designated Author, Reviewer and Approver Rev #

Author Reviewer Approver 6

F. Navarro V. Makodym A. Stavenko S. Sheilds 1 Release of Documents classified as Privileged Intellectual Property (PIP) to any third-party without express approval of the Company President or CEO is prohibited. Documents classified as Holtec Proprietary can be shared with the client if so specified in the contract documents.

Non-confidential documents have no restriction on their distribution. Classification of the confidentiality level of a report is made by a Director level (or above) official in the contract holder Enterprise Unit.

Proprietary Information Proprietary Information is annotated in this document by placing the information in bold square brackets [ ] except for some appendices and supplement as noted below. The annotation of the proprietary information corresponds to the specific reason(s) for claiming the information as proprietary as delineated in the respective Affidavit executed by the owners of the information.

The annotations used are provided as follows:

1) Holtec proprietary information - denoted with 4a,4b superscript, which provides the reference the corresponding subsection of the Holtec Affidavit providing the reason(s)

2) Westinghouse proprietary information - denoted with W superscript. Corresponding reason(s) are delineated in the Westinghouse Affidavit

3) Constellation Energy proprietary information - denoted with C superscript, which provides the reference the corresponding subsection of the Holtec Affidavit providing the reason(s)

Appendices A and B contain miscellaneous details of the Holtec methodology, and Appendices C and D and Supplement 1 contain Constellation Energy, Westinghouse and Holtec proprietary information. Hence these appendices and supplement are considered proprietary in their entirety, without placing any information in brackets.

Project No. 80944 Report No. HI-982094 Page i Holtec International Proprietary Information Holtec Report Number: HI-982094 Report Name: Criticality Evaluation for the Byron/Braidwood Rack Installation Project Holtec Project Number: 80944 Summary of Revisions:

Revision 1 This revision is the result of a rack design change. Boral panels were added to the exterior cell walls in the rack cut-out areas around the fuel elevator and the tool bracket (see Holtec drawing 2337). In the main body of the report, the modifications are indicated in the right-hand margin with revision bars (only page 19 is affected). Additionally, page D-11 was replaced.

Revision 2 This revision corrected some of the discussions about the rack-to-rack gaps and added a couple of MCNP calculations to analyze the rack-to-rack gap. Subsection 7.1.3.1 was added as a result of these changes. Worksheet C.5 was added to show the new MCNP calculations. Revision bars in the right-hand margin indicate changes in text. Revision bars from Revision 1 were removed.

MCNP runs bbr02e, bbr03e, bbr04e, and bbr05e were added to the list in Appendix E.

Revision 3 This revision adds Supplement 1 under a new project number, 2216, to addressing the reactivity effect of missing fuel rods in the Region 2 racks.

Revision 4 This revision consists of editorial changes only. The footnote symbols on pages 26 and 27 were corrected. A typo in section 7.2.1 was rectified. The review and certification log pages for the older revisions were removed.

Revision 5 This revision consists of text added to Section 7 to address the impact of fuel placards. These changes are performed under project 1677.

Revision 6 Previous changes are accepted. Proprietary markings are added.

Project No. 80944 Report No. HI-982094 Page ii Holtec International Proprietary Information Table Of Contents

1.

INTRODUCTION............................................................................................................................................... 1 1.1 DEFINITION AND PURPOSE OF A CALCULATION PACKAGE.............................................................................. 1

2.

METHODOLOGY.............................................................................................................................................. 2

3.

ACCEPTANCE CRITERIA............................................................................................................................... 3

4.

ASSUMPTIONS.................................................................................................................................................. 4

5.

INPUT DATA...................................................................................................................................................... 5 5.1 FUEL ASSEMBLY SPECIFICATION.................................................................................................................... 5 5.2 HOLTEC STORAGE RACK SPECIFICATION....................................................................................................... 5 5.2.1 Region 1 Style........................................................................................................................................ 5 5.2.2 Region 2 Style........................................................................................................................................ 6

6.

COMPUTER CODES......................................................................................................................................... 7

7.

ANALYSIS........................................................................................................................................................... 8 7.1 REGION 1........................................................................................................................................................ 8 7.1.1 Identification of Reference Fuel Assembly............................................................................................. 9 7.1.2 Uncertainties Due to Manufacturing Tolerances.................................................................................. 9 7.1.3 Calculation of Maximum keff.................................................................................................................. 9 7.1.3.1 Water-Gap Spacing Between Racks................................................................................................................. 10 7.1.4 Abnormal and Accident Conditions..................................................................................................... 10 7.1.4.1 Temperature and Water Density Effects.......................................................................................................... 10 7.1.4.2 Eccentric Fuel Assembly Positioning............................................................................................................... 10 7.1.4.3 Dropped Assembly - Horizontal....................................................................................................................... 11 7.1.4.4 Dropped Assembly - Vertical........................................................................................................................... 11 7.1.4.5 Abnormal Location of a Fuel Assembly.......................................................................................................... 12 7.1.4.6 Lateral Rack Movement................................................................................................................................... 12 7.1.5 Summary.............................................................................................................................................. 12 7.2 REGION 2...................................................................................................................................................... 12 7.2.1 Identification of Reference Fuel Assembly........................................................................................... 13 7.2.2 Uncertainties Due to Manufacturing Tolerances................................................................................ 13 7.2.3 Reactivity Effect of Axial Burnup Distribution.................................................................................... 13 7.2.3.1 Axial Burnup Distribution for Assemblies with Axial Blankets...................................................................... 14 7.2.3.2 Axial Burnup Distribution for Assemblies without Axial Blankets................................................................. 15 7.2.4 Uncertainty in Depletion Calculations................................................................................................ 16 7.2.5 Calculation of Maximum keff................................................................................................................ 16 7.2.6 Determination of Burnup Versus Enrichment Curve........................................................................... 16 7.2.7 Abnormal and Accident Conditions..................................................................................................... 17 7.2.7.1 Temperature and Water Density Effects.......................................................................................................... 17 7.2.7.2 Eccentric Fuel Assembly Positioning............................................................................................................... 17 7.2.7.3 Dropped Assembly - Horizontal....................................................................................................................... 18 7.2.7.4 Dropped Assembly - Vertical........................................................................................................................... 18 7.2.7.5 Abnormal Location of a Fuel Assembly.......................................................................................................... 18 7.2.7.5.1 Misloaded Fresh Fuel Assembly............................................................................................................... 18 7.2.7.5.2 Mislocated Fresh Fuel Assembly.............................................................................................................. 19 7.2.7.6 Lateral Rack Movement................................................................................................................................... 19 7.2.8 Summary.............................................................................................................................................. 20 7.3 LONG TERM REACTIVITY CHANGES............................................................................................................. 20

8.

COMPUTER FILES......................................................................................................................................... 21

9.

CONCLUSIONS................................................................................................................................................ 22

Project No. 80944 Report No. HI-982094 Page iii Holtec International Proprietary Information 9.1 REGION 1...................................................................................................................................................... 22 9.2 REGION 2...................................................................................................................................................... 22

10.

REFERENCES.............................................................................................................................................. 23 APPENDIX A: Benchmark Calculations A-1 APPENDIX B: Boral Composition B-1 APPENDIX C: Region 1 PWR Fuel Calculations C-1 APPENDIX D: Region 2 PWR Fuel Calculations D-1 APPENDIX E: List of Computer Runs F-1 Supplement 1: Evaluation of Fuel Assemblies with Missing Fuel Rods for Storage in Region 2 S1-1 Attachment A: HOLTEC APPROVED COMPUTER PROGRAM LIST A1

Project No. 80944 Report No. HI-982094 Page 1 Holtec International Proprietary Information

1. INTRODUCTION This report documents the criticality safety evaluation for the storage of PWR spent nuclear fuel in Holtec Region 1 & 2 style high-density spent fuel storage racks at the Byron and Braidwood nuclear power plants operated by Commonwealth Edison (ComEd). The objective of this analysis is to ensure that the effective neutron multiplication factor (keff) is less than or equal to 0.95 with the storage racks fully loaded with fuel of the highest anticipated reactivity and the pool flooded with unborated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than 0.95 with a 95% probability at a 95% confidence level [1]. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95. The purpose of the present analysis is to confirm the acceptability of the storage rack designs.

1.1 Definition and Purpose of a Calculation Package This Calculation Package has been prepared pursuant to the provisions of Holtec Quality Procedures HQP 3.0 and 3.2, which require that all analyses utilized in support of the design of a safety-related or important-to-safety structure, component, or system be fully documented such that the analyses can be reproduced at any time in the future by a specialist trained in the discipline(s) involved. HQP 3.2 sets down a rigid format structure for the content and organization of Calculation Packages which are intended to create a document which is complete in terms of the exhaustiveness of content. Calculation Packages, however, may lack the narrational smoothness of a Technical Report, and are not intended to serve as a Technical Report.

Accordingly, this Calculation Package is compiled to provide archival information to supplement the material presented in Chapter 4 of the licensing amendment report HI-982083 [2]. The material presented in this Calculation Package is not needed to comprehend the material presented in the above-mentioned Technical Report (which is a self-contained document in full compliance with the USNRC regulations), unless the reader wishes to examine the computational details.

Because of the Calculation Packages function as a repository of all analyses performed on the subject of its scope, this document is typically revised only if an error is discovered in the computations or the equipment design is modified. Additional analyses in the future will be added as numbered supplements to this Package. Each time a supplement is added or the existing material is revised, the revision status of this Package is advanced to the next number and the Table of Contents is amended.

Project No. 80944 Report No. HI-982094 Page 2 Holtec International Proprietary Information

2. METHODOLOGY The principal method for the criticality analysis of the high density storage racks is the three-dimensional Monte Carlo code MCNP4a [3]. MCNP4a is a continuous energy three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory. MCNP4a was selected because it has been extensively used and verified for criticality analyses and has all of the necessary features for this analysis. MCNP4a calculations used continuous energy cross-section data based on ENDF/B-V, as distributed with the code [3]. Independent verification calculations were performed with KENO5a [4], which is a three-dimensional multigroup Monte Carlo code developed at the Oak Ridge National Laboratory. The KENO5a calculations used the 238-group cross-section library, which is based on ENDF/B-V data and is distributed as part of the SCALE-4.3 package [5], in association with the NITAWL-II program [6], which adjusts the uranium-238 cross sections to compensate for resonance self-shielding effects. Benchmark calculations, presented in Appendix A, indicate a bias of [ ]4a,4b with an uncertainty of [ ]4a,4b for KENO5a and [ ]4a,4b for MCNP4a, both evaluated with a 95% probability at the 95% confidence level [1]. The calculations for this analysis utilize the same computer platform and cross-section libraries used for the benchmark calculations discussed in Appendix A.

The convergence of a Monte Carlo criticality problem is sensitive to the following parameters:

(1) number of histories per cycle, (2) the number of cycles skipped before averaging, (3) the total number of cycles and (4) the initial source distribution. The MCNP4a criticality output contains a great deal of useful information that may be used to determine the acceptability of the problem convergence. This information has been used in parametric studies to develop appropriate values for the aforementioned criticality parameters to be used in storage rack criticality calculations.

Based on these studies, a minimum of 5,000 histories were simulated per cycle, a minimum of 20 cycles were skipped before averaging, a minimum of 200 cycles were accumulated, and the initial source was specified as uniform over the fueled regions (assemblies). Further, the output was reviewed to ensure that each calculation achieved acceptable convergence. These parameters represent an acceptable compromise between calculational precision and computational time.

Fuel depletion analyses during core operation were performed with CASMO-4 (using the 40-group cross-section library), a two-dimensional multigroup transport theory code based on capture probabilities [7-9]. Restarting the CASMO-4 calculations in the storage rack geometry at 4 qC yields the two-dimensional infinite multiplication factor (kinf) for the storage rack. Parallel calculations with CASMO-4 for the storage rack at various enrichments enable a reactivity equivalent enrichment (fresh fuel) to be determined that provides the same reactivity in the rack as the depleted fuel. CASMO-4 was also used to determine the small reactivity uncertainties (differential calculations) associated with manufacturing tolerances.

In the geometric models used for the calculations, each fuel rod and its cladding were described explicitly and reflecting boundary conditions were used in the radial direction which has the effect of creating an infinite radial array of storage cells.

Project No. 80944 Report No. HI-982094 Page 3 Holtec International Proprietary Information

3. ACCEPTANCE CRITERIA The high density spent fuel PWR storage racks for ComEd are designed in accordance with the applicable codes and standards listed below. The objective of this evaluation is to show that the effective neutron multiplication factor, keff, is equal to or less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with un-borated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than 0.95 with a 95% probability at a 95% confidence level [1]. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95.

Applicable codes, standard, and regulations or pertinent sections thereof, include the following:

x Code of Federal Regulations, Title 10, Part 50, Appendix A, General Design Criterion 62, Prevention of Criticality in Fuel Storage and Handling.

x USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, Spent Fuel Storage, Rev. 3

- July 1981.

x USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18, 1979.

x L. Kopp, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, NRC Memorandum from L. Kopp to T.

Collins, August 19, 1998.

x USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2 (proposed), December 1981.

x ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.

USNRC guidelines [10] and the applicable ANSI standards specify that the maximum effective multiplication factor, keff, including bias, uncertainties, and calculational statistics, shall be less than or equal to 0.95, with 95% probability at the 95% confidence level. In the present criticality safety evaluation of the ComEd storage racks, the design limit was assumed to be 0.945, which is more conservative than the limit specified in the regulatory guidelines.

Project No. 80944 Report No. HI-982094 Page 4 Holtec International Proprietary Information

4. ASSUMPTIONS To assure the true reactivity will always be less than the calculated reactivity, the following conservative design criteria and assumptions were employed:

1) Moderator is unborated water at a temperature that results in the highest reactivity (4qC, corresponding to the maximum possible moderator density).

2) No soluble poison or control rods are assumed to be present for normal operations.

3) Neutron absorption in minor structural members is neglected, i.e., spacer grids are replaced by water.

4) No Integral Fuel Burnable Absorber (IFBA) rods are assumed to be present in the fuel assemblies.

5) The effective multiplication factor of an infinite radial array of fuel assemblies was used in the analyses, except for the assessment of peripheral effects and certain abnormal/accident conditions where neutron leakage is inherent.

6) Depletion calculations assume conservative operating conditions; highest fuel and moderator temperature and an allowance for the soluble boron concentrations during in-core operation.

Project No. 80944 Report No. HI-982094 Page 5 Holtec International Proprietary Information

5. INPUT DATA 5.1 Fuel Assembly Specification The spent fuel storage racks are designed to accommodate Westinghouse 17x17 OFA, 17x17 Vantage 5, and 17x17 Vantage+ fuel assemblies. The design specifications for these fuel assemblies, which were used for this analysis, are given in Table 5.1 [11,16].

5.2 Holtec Storage Rack Specification The storage cell characteristics were taken from the following Holtec drawings.

x Holtec Drawing 2337 - latest revision x Holtec Drawing 2338 - latest revision x Holtec Drawing 2339 - latest revision x Holtec Drawing 2340 - latest revision x Holtec Drawing 2341 - latest revision The information on these drawings is summarized later in this section.

The tolerance on the box wall stainless steel thickness is taken from the ASME standard for sheet material. Rack fabrication specifies a minimum B-10 loading value for the Boral panels, with a nominal B-10 loading [ ]4a,4b times the minimum loading. The cladding on the Boral, which is nominally [ ]4a,4b of Aluminum is assumed to be homogenized with the B4C and Al mixture in the center of the panel. The Boral composition is listed in Appendix B.

5.2.1 Region 1 Style The Region 1 storage cells are composed of stainless steel boxes separated by a gap with fixed neutron absorber panels, Boral, centered on each side in a 0.110 inch channel. The 0.075 +/-

[ ]4a,4b thick steel walls define the storage cells which have a 8.82 +/- [ ]4a,4b inch nominal inside dimension. A 0.0235 inch stainless steel sheath supports the Boral panel and defines the boundary of the flux-trap water-gap used to augment reactivity control. The cells are located on a lattice spacing of 10.574 +/- [ ]4a,4b inch in one direction and 10.888 +/- [ ]4a,4b inch in the other direction. Stainless steel channels connect the storage cells in a rigid structure and define the flux-trap between the Boral panels, which are 1.337 +/- [ ]4a,4b inch in one direction and 1.651 +/- [ ]4a,4b inch in the other direction. The Boral absorber has a thickness of 0.101 +/- [ ]4a,4b inch and a nominal B-10 areal density of 0.0324 g/cm2 (minimum of 0.0300 g/cm2). The Boral absorber panels are 7.5 +/- [ ]4a,4b inches in width and 147

[ ]4a,4b inches in length. Boral panels are installed on all exterior walls facing other racks, as well as, non-fueled regions, i.e., the pool walls. The minimum gap between neighboring

Project No. 80944 Report No. HI-982094 Page 6 Holtec International Proprietary Information Region 1 style racks and between Region 1 and Region 2 style racks is greater than or equal to 1.75 inches.

Figure 5.1 shows a detailed illustration of the calculational model of the nominal Region 1 spent fuel storage cell, including dimensions. The calculational models consists of a single cell with reflective boundary conditions through the centerline of the water gaps, thus simulating an infinite array of Region 1 storage cells. Figure 5.2 shows the actual MCNP4a calculational model containing the reference 17x17 assembly, as drawn by the two-dimensional plotter in MCNP4a.

5.2.2 Region 2 Style The Region 2 storage cells are composed of stainless steel walls with a single fixed neutron absorber panel, Boral, (attached by a 0.035 stainless steel sheathing) centered on each side in a 0.110 inch channel. Stainless steel boxes are arranged in an alternating pattern such that the connection of the box corners form storage cells between those of the stainless steel boxes. These cells are located on a lattice spacing of 8.97 +/- [ ]4a,4b inch. The 0.075 +/- [ ]4a,4b thick steel walls define a storage cell which has a 8.75 +/- [ ]4a,4b inch nominal inside dimension.

The Boral absorber has a thickness of 0.101 +/- [ ]4a,4b inch and a nominal B-10 areal density of 0.0324 g/cm2 (minimum of 0.0300 g/cm2). The Boral absorber panels are 7.5 +/- [ ]4a,4b inches in width and 147 [ ]4a,4b inches in length. Boral panels are installed on one side of neighboring Region 2 racks. Boral panels are not installed on exterior walls facing non-fueled regions, i.e., the pool walls. The minimum gap between neighboring Region 2 style racks is 0.875 inches, while the minimum gap between Region 1 and Region 2 style racks is 1.75 inches.

Figure 5.3 shows a detailed illustration of the calculational model of the nominal Region 2 spent fuel storage cell, including dimensions. The calculational model consists of a single cell with reflective boundary conditions through the centerline of the composite of materials between the cells, thus simulating an infinite array of Region 2 storage cells. The composite wall thickness shown on Figure 5.3 is half the thickness of the total composite of materials (i.e., [0.075 + 0.035

+ 0.110]/2 = 0.22/2 = 0.110 inches). The thickness of the Boral panels in the model is the half-thickness (0.101/2 = 0.0505 inches). Figure 5.4 shows the actual MCNP4a calculational model containing the reference 17x17 assembly, as drawn by the two-dimensional plotter in MCNP4a

Project No. 80944 Report No. HI-982094 Page 7 Holtec International Proprietary Information

6. COMPUTER CODES The following computer codes were used during this analysis.

x MCNP4a [3] is a three-dimensional continuous energy Monte Carlo code developed at Los Alamos National Laboratory. This code offers the capability of performing full three-dimensional calculations for the loaded storage racks. MCNP4a was run on the PCs at Holtec.

x KENO5a [4] is three-dimensional multigroup Monte Carlo code developed at the Oak Ridge National Laboratory. The KENO5a calculations used a 238-group cross-section library and NITAWL [6] for 238U resonance shielding effects (Nordheim integral treatment). This code offers the capability of performing full three-dimensional calculations for the loaded storage racks. KENO5a was run on the PCs at Holtec.

x CASMO-4 [7-9] is a two-dimensional multigroup transport theory code developed by Studsvik of America. CASMO-4 performs cell criticality calculations and burnup. CASMO-4 has the capability of analytically restarting burned fuel assemblies in the rack configuration.

This code was used to determine the reactivity effects of tolerances and fuel depletion. The CASMO-4 code was run on a DEC 500au workstation at Holtec.

Project No. 80944 Report No. HI-982094 Page 8 Holtec International Proprietary Information

7. ANALYSIS This section describes the calculations that were used to determine the acceptable storage criteria for both the Region 1 and Region 2 style racks. In addition, this section discusses the possible abnormal and accidents conditions.

Unless otherwise stated, all calculations assumed nominal characteristics for the fuel and the fuel storage cells. The effect of the manufacturing tolerances is accounted for with a reactivity adjustment as discussed below.

As discussed in Section 2, MCNP4a was the primary code used in the PWR calculations.

CASMO-4 was used to determine the reactivity effect of tolerances and for depletion calculations. MCNP4a was used for reference cases and to perform calculations which are not possible with CASMO-4 (e.g. eccentric fuel positioning, axial burnup distributions, and fuel misloading). KENO5a was used for independent verification of the reference cases.

Figures 5.2 and 5.4 are pictures of the calculational models used in MCNP4a and CASMO-4.

These pictures were created with the two-dimensional plotter in MCNP4a and clearly indicate the explicit modeling of fuel rods in each fuel assembly. Since CASMO-4 is a two-dimensional code, the fuel assembly hardware above and below the active fuel length is not represented. The three-dimensional MCNP4a and KENO5a models that included axial leakage assumed 30 cm of water above and below the active fuel length. Note that this approach excludes any rack and fuel assembly structural components above or below the active fuel length, including bottom or top nozzles, fuel placards, rack cell wall and poison material, etc. The approach is acceptable because it is conservative to neglect the additional neutron absorbers associated with the rack and absorber materials. Furthermore, the end of the fuel assembly is a high neutron leakage area and thus an area of low importance and therefore, a full density 12 inch water moderator reflector is sufficient.

7.1 Region 1 The goal of the criticality calculations for the PWR Region 1 style racks is to qualify the racks for storage of new unburned fuel assemblies with design specifications as shown in Table 5.1 and a maximum nominal enrichment of 5.0 wt% 235U.

Appendix C contains the results of all of the PWR Region 1 calculations performed for this report. The sheets in Appendix C are printouts of Worksheets from an EXCEL file. In the discussion that follows, these Worksheets will be referenced by their number, which is identified in the top center of each page. A listing of the Worksheets in Appendix C is provided, along with the EXCEL filename, on page C-2 of Appendix C.

Project No. 80944 Report No. HI-982094 Page 9 Holtec International Proprietary Information 7.1.1 Identification of Reference Fuel Assembly In terms of dimensions that are important to reactivity, all of the assembly types that the racks are to accommodate are identical. Therefore, calculations to determine the most reactive assembly type are not necessary. Herein, the reference fuel assembly is referred to as a Westinghouse 17x17 assembly, with dimensions listed in Table 5.1. This assembly, with an initial nominal enrichment of 5.0 wt% 235U was used for all Region 1 calculations.

7.1.2 Uncertainties Due to Manufacturing Tolerances In the calculation of the final kinf, the effect of manufacturing tolerances on reactivity must be included. CASMO-4 was used to perform these calculations. The reference fuel assembly with an initial enrichment of 5.0 wt% 235U was used for these studies. Due to modeling limitations in CASMO-4, the average of the two water gap thicknesses was used. To determine the 'k associated with a specific manufacturing tolerance, the reference kinf was compared to the kinf from a calculation with the tolerance included. All of the 'k values from the various tolerances are statistically combined (square root of the sum of the squares) to determine the final reactivity allowance for manufacturing tolerances. Only the 'k values in the positive direction (increasing reactivity) were used in the statistical combination.

The following is a list of the manufacturing tolerances that were included.

UO2 density:

[ ]W Enrichment:

[ ]W Pitch:

[ ]W Box wall thickness:

[ ]4a,4b Boral width:

[ ]4a,4b B-10 loading in Boral:

[ ]4a,4b Worksheet C.1 in Appendix C shows the kinf from the reference case as compared to the kinf from the cases with the manufacturing tolerances included. The statistical combination of these values is also shown. Because Region 1 is designed for storage of fresh fuel assemblies, all of the individual reactivity allowances were calculated for the reference fresh unburned fuel assembly.

7.1.3 Calculation of Maximum keff Using the calculational model shown in Figure 5.2 and the reference 17x17 fuel assembly, the keff in the Region 1 storage racks has been calculated with both MCNP4a and KENO5a. The determination of the maximum keff, which is based on the MCNP4a calculated keff, the code bias, and the applicable uncertainties, is summarized in Table 7.1. Uncertainties associated with depletion (e.g., depletion uncertainty and axial burnup distribution penalty) are not applicable to Region 1. The reactivity effect of eccentric fuel positioning will be shown to be negative in a later section. Finally, the determination of the maximum keff, based on the KENO5a independent

Project No. 80944 Report No. HI-982094 Page 10 Holtec International Proprietary Information verification calculation, as well as a comparison between the MCNP4a and KENO5a results is shown on Worksheet C.4 in Appendix C.

7.1.3.1 Water-Gap Spacing Between Racks The Boral panels installed on the external surfaces of interfacing Region 1 and Region 2 racks and the minimum spacing between racks, which is 1.75 inches between Region 1 style racks and also 1.75 inches between Region 1 and Region 2 style racks, constitutes a neutron flux-trap between the storage cells of facing racks. The racks are constructed with the base plates extending beyond the edge of the cells which assures that the minimum spacing between storage racks is maintained under all credible conditions. The water-gap flux-trap formed between racks is 1.38 inches which is larger than the 1.337 inch water-gap flux-trap in one direction and smaller than the 1.651 inches in the other direction. Calculations reported on Worksheet C.5 in Appendix C demonstrate that this slight reduction in the water-gap flux-trap between racks compared to the internal water-gap flux-trap (1.38 inches compared to 1.651 inches has a negligible impact

([ ]4a,4b) on the calculated maximum reactivity (Worksheet C.4). Therefore, boundary constraints between rack modules are not necessary or required.

7.1.4 Abnormal and Accident Conditions The effects on reactivity of credible abnormal and accident conditions are examined in this section. None of the abnormal or accident conditions that have been identified as credible will result in exceeding the limiting reactivity (keff d 0.95). The double contingency principal of ANSI N16.1-1975 (and the USNRC letter of April 1978) specifies that it shall require at least two unlikely independent and concurrent events to produce a criticality accident. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

7.1.4.1 Temperature and Water Density Effects The spent fuel pool temperature coefficient of reactivity is negative. Using the temperature of maximum possible water density (4°C), therefore, assures that the true reactivity will always be lower than the calculated value regardless of temperature. Temperature effects on reactivity have been calculated with CASMO-4 and the results are presented on Worksheet C.2 in Appendix C. Introducing voids in the water internal to the storage cell (to simulate boiling) decreased reactivity, as shown on the worksheet. Boiling at the submerged depth of the racks would occur at approximately 122°C.

7.1.4.2 Eccentric Fuel Assembly Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell.

Nevertheless, MCNP4a calculations were made with the fuel assemblies assumed to be in the corner of the storage rack cell (four-assembly cluster at closest approach). These calculations

Project No. 80944 Report No. HI-982094 Page 11 Holtec International Proprietary Information indicated that eccentric fuel positioning results in a decrease in reactivity as shown on Worksheet C.3 in Appendix C. The highest reactivity, therefore, corresponds to the reference design with the fuel assemblies positioned in the center of the storage cells.

7.1.4.3 Dropped Assembly - Horizontal For the case in which a fuel assembly is assumed to be dropped on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the active fuel region of more than 12 inches, which is sufficient to preclude neutron coupling (i.e., an effectively infinite separation). Maximum expected deformation under seismic or accident conditions will not reduce the minimum spacing to less than 12 inches. Consequently, the horizontal fuel assembly drop accident will not result in a significant increase in reactivity. Furthermore, the soluble boron in the spent fuel pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident.

7.1.4.4 Dropped Assembly - Vertical It is also possible to vertically drop an assembly into a location occupied by another assembly.

Such a vertical impact would at most cause a small compression of the stored assembly, reducing the water-to-fuel ratio and thereby reducing reactivity. In addition the distance between the active fuel regions of both assemblies will be more than sufficient to ensure no significant neutron interaction between the two assemblies.

Structural analysis [12] has shown that dropping an assembly into an unoccupied cell could result in a localized deformation of the baseplate of the rack. The resultant effect would be the lowering of a single fuel assembly by the amount of the deformation. This could potentially result in the active fuel height of that assembly no longer being completely covered by the Boral.

The immediate eight surrounding fuel cells could also be affected. However, the amount of deformation for these cells would be considerably less. Structural analysis has shown that the amount of localized deformation will not exceed two inches [12]. The reactivity consequence of this situation was investigated and is summarized on Worksheet C.3 in Appendix C. For simplicity in modeling, the calculation conservatively assumed an infinite array of assemblies in this damaged condition, and demonstrated the reactivity effect to be negligible. Since this is a localized event (nine storage cells at most) the actual reactivity effect will be even less than the calculated value.

Furthermore, the soluble boron in the spent fuel pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident. Consequently, a dropped fuel bundle resulting in a localized deformation of less than two inches will have a negligible impact on reactivity.

This analysis with two inches of exposed fuel in all cells will also bound any condition related to elongation of the fuel during irradiation.

Project No. 80944 Report No. HI-982094 Page 12 Holtec International Proprietary Information 7.1.4.5 Abnormal Location of a Fuel Assembly The Region 1 racks are qualified for the storage of all fresh unburned fuel assemblies with the maximum permissible enrichment (5.0 wt% 235U). Therefore the abnormal location of a fuel assembly in Region 1 is of no concern.

7.1.4.6 Lateral Rack Movement Lateral motion of the storage racks under seismic conditions could potentially alter the spacing between Region 1 racks. However, the baseplate extensions prohibit this spacing from being reduced beyond the minimum 1.75 inches specified and analyzed in Section 7.1.3.1.

Furthermore, soluble poison would assure that a reactivity less that the design limitation is maintained under seismic conditions. Consequently, there will be no positive effect on reactivity as a result of lateral rack movement.

7.1.5 Summary Calculations have been performed to qualify the Region 1 racks for storage of new unburned fuel assemblies with a maximum nominal enrichment of 5.0 wt% 235U. The criticality analyses for Region 1 of the spent fuel storage pool are summarized in Table 7.1, and demonstrate that for the defined acceptance criteria, the maximum keff is less than 0.945. An independent calculation with the KENO5a code provides confirmation of the validity of the reference MCNP4a calculations.

Worksheet C.4 in Appendix C summarizes the MCNP4a and KENO5a comparison.

7.2 Region 2 The goal of the criticality calculations for the PWR Region 2 style racks is to qualify the racks for storage of fuel assemblies with design specifications as shown in Table 5.1 and a maximum nominal initial enrichment of 5.0 wt% 235U that have accumulated a minimum burnup of 40 GWD/MTU. Additionally, the criticality calculations are to determine the initial enrichment and burnup combinations required for the storage of fuel assemblies with nominal initial enrichments up to 5.0 wt% 235U.

Appendix D contains the results of all of the PWR Region 2 calculations performed for this report. The sheets in Appendix D are printouts of Worksheets from an EXCEL file. In the discussion that follows, these Worksheets will be referenced by their number, which is identified in the top center of each page. A listing of the Worksheets in Appendix D is provided, along with the EXCEL filename, on page D-2 of Appendix D.

Project No. 80944 Report No. HI-982094 Page 13 Holtec International Proprietary Information 7.2.1 Identification of Reference Fuel Assembly In terms of dimensions that are important to reactivity, all of the assembly types that the racks are to accommodate are identical. Therefore, calculations to determine the most reactive assembly type are not necessary. Herein, the reference fuel assembly is referred to as a Westinghouse 17x17 assembly, with dimensions listed in Table 5.1. This assembly, with an initial nominal enrichment of 5.0 wt% 235U was used for all Region 2 calculations.

7.2.2 Uncertainties Due to Manufacturing Tolerances In the calculation of the final kinf, the effect of manufacturing tolerances on reactivity must be included. CASMO-4 was used to perform these calculations. The reference fuel assembly with an initial enrichment of 5.0 wt% 235U was used for these studies. To determine the 'k associated with a specific manufacturing tolerance, the reference kinf was compared to the kinf from a calculation with the tolerance included. All of the 'k values from the various tolerances are statistically combined (square root of the sum of the squares) to determine the final reactivity allowance for manufacturing tolerances. Only the 'k values in the positive direction (increasing reactivity) were used in the statistical combination.

The following is a list of the manufacturing tolerances that were included.

UO2 density:

[ ]W Enrichment:

[ ]W Pitch:

[ ]W Box wall thickness:

[ ]4a,4b Boral width:

[ ]4a,4b B-10 loading in Boral:

[ ]4a,4b Worksheet D.1 in Appendix D shows the kinf from the reference case as compared to the kinf from the cases with the manufacturing tolerances included. The statistical combination of these values is also shown. All of the individual reactivity allowances were calculated for the reference fuel assembly at zero burnup and at burnups enveloping the criteria for storage (i.e., 35, 40, and 45 GWD/MTU). Conservatively, the largest total uncertainty from either the fresh or burned condition was used.

There is also a reactivity allowance for an uncertainty in the burnup calculations which must be included in the final statistical combination of the reactivity allowances. This is discussed in a later section.

7.2.3 Reactivity Effect of Axial Burnup Distribution Initially, fuel loaded into the reactor will burn with a slightly skewed cosine power distribution.

As burnup progresses, the burnup distribution will tend to flatten, becoming more highly burned

Project No. 80944 Report No. HI-982094 Page 14 Holtec International Proprietary Information in the central regions than in the upper and lower ends. At high burnup, the more reactive fuel near the ends of the fuel assembly (less than average burnup) occurs in regions of lower reactivity worth due to neutron leakage. Consequently, it would be expected that over most of the burnup history, distributed burnup fuel assemblies would exhibit a slightly lower reactivity than that calculated for the average burnup. As burnup progresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of large regions of significantly reduced burnup.

Generic analytic results of the axial burnup effect for assemblies without axial blankets have been provided by Turner [13] based upon calculated and measured axial burnup distributions.

These analyses confirm the minor and generally negative reactivity effect of the axially distributed burnup, becoming positive at burnups greater than about 30 GWD/MTU. The trends observed [13] suggest the possibility of a small positive reactivity effect above 30 GWD/MTU increasing to slightly over 1% 'k at 40 GWD/MTU. Since the required burnup for fuel of 5 wt%

235U initial enrichment is greater than 30 GWD/MTU, the reactivity effect of the axially distributed burnup must be considered.

The Byron and Braidwood plants possess fuel assemblies with and without low enrichment axial blankets on the ends, which effect the axial burnup distribution. ComEd has supplied axial burnup distribution data [14,15] for both types of assembly designs and the effect of the axial burnup distribution of both types of assembly designs is examined in this section.

7.2.3.1 Axial Burnup Distribution for Assemblies with Axial Blankets The fuel assemblies with axial blankets have low enrichment ([ ]W) regions that are 6 inches in length at either end (top and bottom) of the active fuel region [16].

ComEd provided four axial burnup distributions for assemblies with axial blankets that have accumulated assembly-average burnups between [ ]C [14]. These data are listed on Worksheet D.5 in Appendix D. The four burnup distributions were averaged to generate a representative axial burnup distribution, with an assembly average burnup of [

]C. This distribution was then normalized to assembly-average burnups of 30.0 and 40.0 GWD/MTU to determine the effect of the axial distribution in burnup at these burnups. The case with 40.0 GWD/MTU corresponds to the required burnup for 5.0 wt% enriched fuel.

Worksheet D.5 in Appendix D shows the calculation of the equivalent enrichments for the 24 axial zones to represent the distributed axial burnup (for assembly-average burnups of 30.0 and 40.0 GWD/MTU). Worksheet D.3 in Appendix D shows the burnup versus kinf data and the curve fits for fuel with initial enrichments of 5.0 wt% and [ ]W (used for the axial blankets). Worksheet D.4 in Appendix D presents the kinf versus enrichment data and the associated curve fits. The curve fits developed in these worksheets were used to calculate the equivalent enrichments for the 24 axial zones to represent the axial burnup distribution.

In order to determine the reactivity effect of the axial burnup distribution for these assemblies, two three-dimensional MCNP4a calculations were performed for each burnup considered (30

Project No. 80944 Report No. HI-982094 Page 15 Holtec International Proprietary Information and 40 GWD/MTU). The first run used an average burnup distribution over the entire axial length. The second run used a distributed axial burnup distribution represented by 24 axial zones (using equivalent enrichments). The keff values from these runs were compared to determine the reactivity effect at the analyzed burnup.

The results of comparing the keff from the calculation using the axial burnup distribution and the calculation using an assembly-average axial burnup are provided on Worksheet D.7 in Appendix D. The results show that using the distributed axial burnup distribution, as opposed to an assembly-average burnup distribution, for assemblies with axial blankets results in a lower keff. The presence of low enrichment regions (axial blankets) at either end of the fuel assembly precludes the existence of high reactivity regions at the ends, which are present in assemblies without axial blankets. Thus the effect of the axial burnup distribution is to reduce the reactivity.

7.2.3.2 Axial Burnup Distribution for Assemblies without Axial Blankets The fuel assemblies without axial blankets have uniform enrichment over the entire active fuel length. ComEd provided an axial burnup distribution for an assembly without axial blankets that has accumulated an assembly-average burnup of [ ]C [15]. These data are listed on Worksheet D.6 in Appendix D. The distribution was normalized to assembly-average burnups of 30.0 and 40.0 GWD/MTU to determine the effect of the axial distribution in burnup at these burnups. The case with 40.0 GWD/MTU corresponds to the required burnup for 5.0 wt%

enriched fuel. Worksheet D.6 in Appendix D shows the calculation of the equivalent enrichments for the 24 axial zones to represent the distributed axial burnup (for assembly-average burnups of 30.0 and 40.0 GWD/MTU). Worksheet D.3 in Appendix D shows the burnup versus kinf data and the curve fits for fuel with an initial enrichment of 5.0 wt%. Worksheet D.4 in Appendix D presents the kinf versus enrichment data and the associated curve fits. The curve fits developed in these worksheets were used to calculate the equivalent enrichments for the 24 axial zones to represent the axial burnup distribution.

In order to determine the reactivity effect of the axial burnup distribution for these assemblies, two three-dimensional MCNP4a calculations were performed for each burnup considered (30 and 40 GWD/MTU). The first run used an average burnup distribution over the entire axial length. The second run used a distributed axial burnup distribution represented by 24 axial zones (using equivalent enrichments). The keff values from these runs were compared to determine the reactivity effect at the analyzed burnup.

The results of comparing the keff from the calculation using the axial burnup distribution and the calculation using an assembly-average axial burnup are provided on Worksheet D.7 presented in Appendix D. The results show that using the distributed axial burnup distribution, as opposed to an average burnup distribution, for assemblies with an assembly-average burnup of 40 GWD/MTU without axial blankets results in a higher keff. This increase in reactivity can be attributed to the presence of high reactivity regions at the ends (low burnup). Consistent with the referenced generic studies [13], the results show that using the distributed axial burnup distribution for assemblies with an assembly-average burnup of 30 GWD/MTU results in a slight

Project No. 80944 Report No. HI-982094 Page 16 Holtec International Proprietary Information reduction in keff. Conservatively, the difference between the two calculations at 40 GWD/MTU is used as a positive reactivity bias for burnups in the 30 to 40 GWD/MTU range.

7.2.4 Uncertainty in Depletion Calculations CASMO-4 was used to perform the depletion calculations. Since critical experiment data with spent fuel is not available for determining the uncertainty in burnup-dependent reactivity calculations, an allowance for uncertainty in reactivity was assigned based upon other considerations. Assuming the uncertainty in depletion calculations is less than 5% of the total reactivity decrement, a burnup dependent uncertainty in reactivity for burnup calculations may be assigned [10]. The uncertainty values are calculated from the kinf versus burnup data for 5.0 wt% 235U initial enrichment fuel provided on Worksheet D.3 in Appendix D. This allowance is statistically combined with the other reactivity allowances in the determination of the maximum keff.

7.2.5 Calculation of Maximum keff Using the calculational model shown in Figure 5.4 and the reference 17x17 fuel assembly with initial enrichment of 5.0 wt% 235U which has accumulated a burnup of 40 GWD/MTU, the keff in the Region 2 storage racks has been calculated with both MCNP4a and KENO5a. The determination of the maximum keff, which is based on the MCNP4a calculated keff, the code bias, and the applicable uncertainties, is summarized in Table 7.2. The reactivity effect of eccentric fuel positioning will be shown to be negative in a later section. Finally, the determination of the maximum keff, based on the KENO5a independent verification calculation, as well as a comparison between the MCNP4a and KENO5a results is shown on Worksheet D.8 in Appendix D.

7.2.6 Determination of Burnup Versus Enrichment Curve Since MCNP4a can not perform depletion it is necessary to use a fresh fuel assembly in the MCNP4a calculations with a 235U enrichment that results in a reactivity equivalent to the desired burnup. CASMO-4 was used to derive this equivalent enrichment. First, a curve is fit to the burnup versus kinf data and then a curve is fit to the enrichment versus kinf data. The equivalent enrichment is then determined by choosing the desired burnup and finding the associated kinf.

The kinf is then used to determine an associated enrichment, the desired equivalent enrichment. A fresh assembly with this enrichment has the same kinf as an assembly with the desired burnup.

Worksheet D.3 in Appendix D shows the burnup versus kinf data and the curve fits. Worksheet D.4 in Appendix D presents the kinf versus enrichment data and the associated curve fits. The burnup versus enrichment curve was calculated by determining the desired kinf and then finding the associated burnup from the kinf versus burnup data taken from CASMO-4. Linear interpolation between designated burnup values was used to determine the specific burnup for

Project No. 80944 Report No. HI-982094 Page 17 Holtec International Proprietary Information the target kinf. The target kinf was based on the calculated kinf value for a fuel assembly with initial nominal enrichment of 5.0 wt% 235U that has achieved a burnup of 40 GWD/MTU.

Worksheet D.9 in Appendix D show the burnups that were calculated for each enrichment that result in the target kinf value. The final burnups versus enrichments were fitted to a curve and this curve was adjusted such that all calculated burnups are on or below the curve. The equation of this bounding curve and a comparison of the calculated burnups to the burnups from the bounding curve-fit are shown on Worksheet D.9. The final calculated burnup versus enrichment data are presented in Figure 7.1, which shows that a minimum burnup of 40 GWD/MTU is required for 5.0 r [ ]W wt% 235U initial enrichment fuel.

7.2.7 Abnormal and Accident Conditions The effects on reactivity of credible abnormal and accident conditions are examined in this section. None of the abnormal or accident conditions that have been identified as credible will result in exceeding the limiting reactivity (keff d 0.95). The double contingency principal of ANSI N16.1-1975 (and the USNRC letter of April 1978) specifies that it shall require at least two unlikely independent and concurrent events to produce a criticality accident. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

7.2.7.1 Temperature and Water Density Effects The spent fuel pool temperature coefficient of reactivity is negative. Using the temperature of maximum possible water density (4°C), therefore, assures that the true reactivity will always be lower than the calculated value regardless of temperature. Temperature effects on reactivity have been calculated with CASMO-4 for the reference fuel assembly at zero burnup and at burnups enveloping the criteria for storage (i.e., 35, 40, and 45 GWD/MTU). The results are presented on Worksheet D.2 in Appendix D. Introducing voids in the water internal to the storage cell (to simulate boiling) decreased reactivity, as shown on the worksheet. Boiling at the submerged depth of the racks would occur at approximately 122°C.

7.2.7.2 Eccentric Fuel Assembly Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell.

Nevertheless, MCNP4a calculations were made with the fuel assemblies assumed to be in the corner of the storage rack cell (four-assembly cluster at closest approach). These calculations indicated that eccentric fuel positioning results in a decrease in reactivity as shown on Worksheet D.7 in Appendix D. The highest reactivity, therefore, corresponds to the reference design with the fuel assemblies positioned in the center of the storage cells.

Project No. 80944 Report No. HI-982094 Page 18 Holtec International Proprietary Information 7.2.7.3 Dropped Assembly - Horizontal For the case in which a fuel assembly is assumed to be dropped on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the active fuel region of more than 12 inches, which is sufficient to preclude neutron coupling (i.e., an effectively infinite separation). Maximum expected deformation under seismic or accident conditions will not reduce the minimum spacing to less than 12 inches. Consequently, the horizontal fuel assembly drop accident will not result in a significant increase in reactivity. Furthermore, the soluble boron in the spent fuel pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident.

7.2.7.4 Dropped Assembly - Vertical It is also possible to vertically drop an assembly into a location occupied by another assembly.

Such a vertical impact would at most cause a small compression of the stored assembly, reducing the water-to-fuel ratio and thereby reducing reactivity. In addition the distance between the active fuel regions of both assemblies will be more than sufficient to ensure no neutron interaction between the two assemblies.

Structural analysis [12] has shown that dropping an assembly into an unoccupied cell could result in a localized deformation of the baseplate of the rack. The resultant effect would be the lowering of a single fuel assembly by the amount of the deformation. This could potentially result in the active fuel height of that assembly no longer being completely covered by the Boral.

The immediate eight surrounding fuel cells could also be affected. However, the amount of deformation for these cells would be considerably less. Structural analysis has shown that the amount of localized deformation will not exceed two inches [12]. The reactivity consequence of this situation was investigated and is summarized on Worksheet D.7 in Appendix D. For simplicity in modeling, the calculation conservatively assumed an infinite array of assemblies in this damaged condition, and demonstrated the reactivity effect to be negligible. Since this is a localized event (nine storage cells at most) the actual reactivity effect will be even less than the calculated value.

Furthermore, the soluble boron in the spent fuel pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident. Consequently, a dropped fuel bundle will have a negligible impact on reactivity.

This analysis with two inches of exposed fuel in all cells will also bound any condition related to elongation of the fuel during irradiation.

7.2.7.5 Abnormal Location of a Fuel Assembly 7.2.7.5.1 Misloaded Fresh Fuel Assembly The misplacement of a fresh unburned fuel assembly could, in the absence of soluble poison, result in exceeding the regulatory limit (keff of 0.95). This could possibly occur if a fresh fuel assembly of

Project No. 80944 Report No. HI-982094 Page 19 Holtec International Proprietary Information the highest permissible enrichment (5.0 wt%) were to be inadvertently misloaded into a Region 2 storage cell intended for burned fuel. The reactivity consequence of this situation was investigated and is summarized on Worksheet D.7 in Appendix D.

The calculational models for the Region 2 misplaced assembly situation consist of 5x5 arrays of assemblies with reflective boundary conditions on all four sides. Comparison of calculational results from these 5x5 models with and without a misplaced assembly demonstrate that, in the absence of soluble boron, the design basis limit of 0.945 is exceeded. Additional calculations were performed to demonstrate that a soluble boron concentration of approximately 212 ppm is required to assure that the 0.945 limit is not exceeded. Thus, the Technical Specifications must require a minimum of 212 ppm of soluble boron in the spent fuel pool to assure that the limiting keff of 0.945 is not exceeded during this accident condition.

7.2.7.5.2 Mislocated Fresh Fuel Assembly The mislocation of a fresh unburned fuel assembly could, in the absence of soluble poison, result in exceeding the regulatory limit (keff of 0.95). This could possibly occur if a fresh fuel assembly of the highest permissible enrichment (5.0 wt%) were to be accidentally mislocated outside of a storage rack adjacent to other fuel assemblies. There are two areas in the pool layout in which such an accident condition could be postulated to occur. The area around the fuel elevator and the area around the tool bracket have sufficient unoccupied space to house a mislocated fuel assembly. The exterior racks walls in these two areas have attached Boral panels.

The worst possible case would be an assembly mislocated in a corner. The reactivity consequence of this situation was investigated and is summarized on Worksheet D.7 in Appendix D.

Comparison of calculational results with and without a mislocated assembly demonstrate that, in the absence of soluble boron, the design basis limit of 0.945 is exceeded. Additional calculations were performed to demonstrate that a soluble boron concentration of approximately 120 ppm is required to assure that the 0.945 limit is not exceeded. Thus, the Technical Specifications must require a minimum of 120 ppm of soluble boron in the spent fuel pool to assure that the limiting keff of 0.945 is not exceeded during this accident condition.

7.2.7.6 Lateral Rack Movement Region 2 storage cells do not use a flux-trap, and thus, the calculated maximum reactivity does not rely on spacing between racks. Nevertheless, the minimum water gap between Region 2 racks (0.875 inches, as limited by the base plate extensions) and the Boral panels, which are installed on one exterior wall of neighboring Region 2 racks, assure that the reactivity is always less than the design limitation. Furthermore, soluble poison would assure that a reactivity less than the design limitation is maintained under seismic conditions. Consequently, there will be no positive effect on reactivity as a result of lateral rack movement.

Project No. 80944 Report No. HI-982094 Page 20 Holtec International Proprietary Information 7.2.8 Summary Calculations have been performed to qualify the Region 2 racks for storage of fuel assemblies with a maximum nominal initial enrichment of 5.0 wt% 235U which have accumulated a minimum burnup of 40.0 GWD/MTU or fuel of initial enrichment and burnup combinations within the acceptable domain depicted in Figure 7.1 The criticality analyses for Region 2 of the spent fuel storage pool are summarized in Table 7.2, and demonstrate that for the defined acceptance criteria, the maximum keff is less than 0.945. An independent calculation with the KENO5a code provides confirmation of the validity of the reference MCNP4a calculations. Worksheet D.8 in Appendix D summarizes the MCNP4a and KENO5a comparison.

7.3 Long Term Reactivity Changes At reactor shutdown, the reactivity of the fuel initially decreases due to the growth of Xe-135.

Subsequently, the Xenon decays and the reactivity increases to a maximum at several hundred hours when the Xenon is gone. Over the next 30 years, the reactivity continuously decreases due primarily to Pu-241 decay and Americium growth. At lower burnup, the reactivity decrease will be less pronounced since less Pu-241 would have been produced. No credit is taken for this long-term decrease in reactivity other than to indicate additional and increasing conservatism in the design criticality analysis.

Project No. 80944 Report No. HI-982094 Page 21 Holtec International Proprietary Information

8. COMPUTER FILES A list of the file names and a brief description of the calculations that were performed for this analysis is provided in Appendix E. All related computer files are stored on the computer server at the Holtec International office in Marlton, New Jersey. The files are stored in the following directory: F:\\PROJECTS\\980944\\JWAGNER.

Project No. 80944 Report No. HI-982094 Page 22 Holtec International Proprietary Information

9. CONCLUSIONS This report documents the criticality analysis for the storage of PWR spent nuclear fuel in Holtec Region 1 & 2 style high-density spent fuel storage racks at the Byron and Braidwood nuclear power plants operated by Commonwealth Edison (ComEd), which supports Chapter 4 of the licensing amendment report HI-982083 [2]. The analysis demonstrates that for the defined acceptance criteria that are summarized below, the effective neutron multiplication factor (keff) is less than or equal to 0.945 with a 95% probability at a 95% confidence level. Further, the reactivity effects of abnormal and accident conditions have been evaluated to assure that under credible abnormal and accident conditions, the reactivity will not exceed 0.945.

9.1 Region 1 Calculations have been performed to qualify the Region 1 racks for storage of fresh unburned fuel assemblies with a maximum nominal enrichment of 5.0 wt% 235U. The criticality analyses for Region 1 of the spent fuel storage pool are summarized in Table 7.1, and demonstrate the maximum keff, which includes bias and all applicable uncertainties, is less than 0.945.

9.2 Region 2 Calculations have been performed to qualify the Region 2 racks for storage of fuel assemblies with a maximum nominal initial enrichment of 5.0 wt% 235U which have accumulated a minimum burnup of 40.0 GWD/MTU or fuel of initial enrichment and burnup combinations within the acceptable domain depicted in Figure 7.1. The criticality analyses for Region 2 of the spent fuel storage pool are summarized in Table 7.2, and demonstrate that for the defined acceptance criteria, the maximum keff, which includes bias and all applicable uncertainties, is less than 0.945.

The calculated maximum reactivity in Region 2 includes the reactivity effect of the axial distribution in burnup and provides an additional margin of uncertainty for the depletion calculations. For convenience, the minimum (limiting) burnup data in Figure 7.1 may be described as a function of the nominal initial enrichment, E, in wt% 235U by a bounding polynomial expression as follows:

[ ]C, where B is the minimum burnup in GWD/MTU and E is the initial enrichment in wt% 235U (for initial enrichments from 2.0 to 5.0 wt% 235U). Fuel assemblies with enrichments less than 2.0 wt% 235U will conservatively be required to meet the burnup requirements of 2.0 wt% 235U assemblies as shown in Figure 7.1. The actual data points on Figure 7.1 are listed in bold at the end of Worksheet D.9 in Appendix D. Linear interpolation between these data points is also an acceptable option.

Project No. 80944 Report No. HI-982094 Page 23 Holtec International Proprietary Information

10. REFERENCES

1. M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

2. Licensing Report for Byron/Braidwood Nuclear Stations, Holtec International Report HI-982083, Project No. 80944, 1998.

3. J.F. Briesmeister, Editor, MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A, LA-12625, Los Alamos National Laboratory (1993).

4. L.M. Petrie and N.F. Landers, KENO Va - An Improved Monte Carlo Criticality Program with Supergrouping, Volume 2, Section F11 from SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation NUREG/CR-0200, Rev. 4, January 1990.

5. SCALE 4.3: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluations, NUREG-CR-0200, Rev. 5, Oak Ridge National Laboratory (1995).

6. N.M. Greene, L.M. Petrie and R.M. Westfall, NITAWL-II: Scale System Module for Performing Shielding and Working Library Production, Volume 1, Section F1 from SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation NUREG/CR-0200, Rev. 4, January 1990.

7. M. Edenius, K. Ekberg, B.H. Forssén, and D. Knott, CASMO-4 A Fuel Assembly Burnup Program Users Manual, Studsvik/SOA-95/1, Studsvik of America, Inc. and Studsvik Core Analysis AB (proprietary).

8. D. Knott, CASMO-4 Benchmark Against Critical Experiments, SOA-94/13, Studsvik of America, Inc., (proprietary).

9. D. Knott, CASMO-4 Benchmark Against MCNP, SOA-94/12, Studsvik of America, Inc., (proprietary).

10. L.I. Kopp, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, NRC Memorandum from L. Kopp to T.

Collins, August 19, 1998.

11. High Density Spent Fuel Racks at Byron and Braidwood Nuclear Stations, issued by Commonwealth Edison Company, Specification NEC-97-7046.

12. Analysis of the Mechanical Accidents for Byron/Braidwood Nuclear Station, Holtec International Report HI-982086, Project No. 80944, 1998.

Project No. 80944 Report No. HI-982094 Page 24 Holtec International Proprietary Information

13. S.E. Turner, "Uncertainty Analysis - Burnup Distributions", presented at the DOE/SANDIA Technical Meeting on Fuel Burnup Credit, Special Session, ANS/ENS Conference, Washington, D.C., November 2, 1988.

14. Nuclear Fuel Services Department Nuclear Design Information Transmittal, NFM9800226, dated 11/4/98.

15. Nuclear Fuel Services Department Nuclear Design Information Transmittal, NFM9800222, dated 11/3/98.

16. Nuclear Fuel Services Department Nuclear Design Information Transmittal, NFM9800218, dated 10/28/98.

Project No. 80944 Report No. HI-982094 Page 25 Holtec International Proprietary Information Table 5.1 PWR Fuel Assembly Specifications Fuel Rod Data Assembly type Westinghouse OFA / Vantage 5 / Vantage+

Fuel pellet outside diameter, in.

0.3088 Cladding thickness, in.

0.0225 Cladding outside diameter, in.

0.360 Cladding material Zr Maximum stack density, g/cc

[ ]W Maximum enrichment, wt% 235U 5.00 r [ ]W Fuel Assembly Data Fuel rod array 17 x 17 Number of fuel rods 264 Fuel rod pitch, in.

[ ]W Number of control rod guide and instrument thimbles 25 Thimble outside diameter, in.

[ ]W Thimble thickness, in.

[ ]W Active fuel Length, in.

144

Project No. 80944 Report No. HI-982094 Page 26 Holtec International Proprietary Information Table 7.1 Summary of the Criticality Safety Analyses for Region 1 Storage Arrangement Unrestricted Design Basis Burnup at 5.0 wt% 235U 0

Uncertainties Bias Uncertainty (95%/95%)

[ ]4a,4b Calculational Statistics 1 (95%/95%, 2.0uV)

[ ]C Depletion Uncertainty N/A Fuel Eccentricity negative Manufacturing Tolerances

[ ]C Statistical Combination of Uncertainties 2

[ ]C Reference keff (MCNP4a)

[ ]C Total Uncertainty (above)

[ ]C Axial Burnup Distribution N/A Calculational Bias (see Appendix A)

[ ]4a,4b Maximum keff 0.9422 3 Regulatory Limiting keff 0.9500 1 The value used for the MCNP4a (or KENO5a) statistical uncertainty is 2.0 times the estimated standard deviation. Each final k value calculated by MCNP4a (or KENO5a) is the result of averaging a minimum of 200 cycle k values, and thus, is based on a minimum sample size of 200. The K multiplier, for a one-sided statistical tolerance with 95% probability at the 95% confidence level, corresponding to a sample size of 200, is 1.84.

However, for this analysis a value of 2.0 was assumed for the K multiplier, which is larger (more conservative) than the value corresponding to a sample size of 200.

2 Square root of the sum of the squares.

3 KENO5a verification calculation resulted in a maximum keff of 0.9429.

Project No. 80944 Report No. HI-982094 Page 27 Holtec International Proprietary Information Table 7.2 Summary of the Criticality Safety Analyses for Region 2 Design Basis Burnup at 5.0 wt% 235U 40.0 GWD/MTU Uncertainties Bias Uncertainty (95%/95%)

[ ]4a,4b Calculational Statistics 4 (95%/95%, 2.0uV)

[ ]C Depletion Uncertainty

[ ]C Fuel Eccentricity negative Manufacturing Tolerances

[ ]C Statistical Combination of Uncertainties 5

[ ]C Reference keff (MCNP4a)

[ ]C Total Uncertainty (above)

[ ]C Axial Burnup Distribution

[ ]C Calculational Bias (see Appendix A)

[ ]4a,4b Maximum keff 0.9377 6 Regulatory Limiting keff 0.9500 4 The value used for the MCNP4a (or KENO5a) statistical uncertainty is 2.0 times the estimated standard deviation. Each final k value calculated by MCNP4a (or KENO5a) is the result of averaging a minimum of 200 cycle k values, and thus, is based on a minimum sample size of 200. The K multiplier, for a one-sided statistical tolerance with 95% probability at the 95% confidence level, corresponding to a sample size of 200, is 1.84.

However, for this analysis a value of 2.0 was assumed for the K multiplier, which is larger (more conservative) than the value corresponding to a sample size of 200.

5 Square root of the sum of the squares.

6 KENO5a verification calculation resulted in a maximum keff of 0.9370.

4a,4b Figure 5.1:

A Cross-Sectional View of the Calculational Model Used for the Region 1 Rack Analysis (NOT TO SCALE).

Project No. 80944 Report No. HI-982094 Page 28

Project No. 80944 Report No. HI-982094 Page 29 Holtec International Proprietary Information Figure 5.2:

A Two-Dimensional Representation of the Actual Calculational Model Used for the Region 1 Rack Analysis. This Figure was Drawn (To Scale) with the Two-Dimensional Plotter in MCNP4a.

4a,4b Figure 5.3:

A Cross-Sectional View of the Calculational Model Used for the Region 2 Rack Analysis (NOT TO SCALE).

Project No. 80944 Report No. HI-982094 Page 30

Project No. 80944 Report No. HI-982094 Page 31 Holtec International Proprietary Information Figure 5.4:

A Two-Dimensional Representation of the Actual Calculational Model Used for the Region 2 Rack Analysis. This Figure was Drawn (To Scale) with the Two-Dimensional Plotter in MCNP4a.

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Project No. 80944 Report No. HI-982094 Page 32 Holtec International Proprietary Information Figure 7.1:

Minimum Required Fuel Assembly Burnup as a Function of Nominal Initial Enrichment to Permit Storage in Region 2 (Fuel assemblies with enrichments less than 2.0 wt% 235U will conservatively be required to meet the burnup requirements of 2.0 wt% 235U assemblies) 0 5

10 15 20 25 30 35 40 45 2

2.5 3

3.5 4

4.5 5

Initial Fuel Enrichment (wt% U-235)

ACCEPTABLE DOMAIN bounding curve actual data points

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Project No. 80944 Report No. HI-982094 Page A-1 Holtec International Proprietary Information Appendix A Benchmark Calculations (total number of pages: 26 including this page)

Note: because this appendix was taken from a different report, the next page is labeled Appendix 4A, Page 1.

Proprietary Information This appendix contains miscellaneous details of Holtec methodology, hence it is considered proprietary in its entirely.

Project No. 80944 Report No. HI-982094 Page B-1 Holtec International Proprietary Information Appendix B Boral Compositions (total number of pages: 2 including this page)

Proprietary Information This appendix contains miscellaneous details of Holtec methodology, hence it is considered proprietary in its entirely.

Project No. 80944 Report No. HI-982094 Page C-1 Holtec International Proprietary Information Appendix C Region 1 PWR Fuel Calculations (total number of pages: 6 including this page)

Proprietary Information This appendix contains miscellaneous details of Constellation Energy, Westinghouse and Holtec proprietary information, hence it is considered proprietary in its entirely.

Project No. 80944 Report No. HI-982094 Page D-1 Holtec International Proprietary Information Appendix D Region 2 PWR Fuel Calculations (total number of pages: 15 including this page)

Proprietary Information This appendix contains miscellaneous details of Constellation Energy, Westinghouse and Holtec proprietary information, hence it is considered proprietary in its entirely.

Project No. 80944 Report No. HI-982094 Page E-1 Holtec International Proprietary Information Appendix E List of Computer Runs (total number of pages: 3 including this one)

Project:

80944 Report:

HI-982094 Input ID Code bbrlab CASMO-3 bbrll0 MCNP4a bbr112 MCNP4a bbr140 MCNP4a bbr141 MCNP4a bbr142 MCNP4a bbr143 MCNP4a bbr144 MCNP4a kbbrla KENO5a bbr02e MCNP4a bbr03e MCNP4a bbr04e MCNP4a bbr05e MCNP4a bbr2t01 CASMO-4 bbr2t02 CASMO-4 bbr2t03 CASMO-4 bbr2t04 CASMO-4 bbr2t05 CASMO-4 bbr2t06 CASMO-4 bbr2d01 CASMO-4 bbr2d02 CASMO-4 bbr2d03 CASMO-4 bbr2d04 CASMO-4 bbr2d05 CASMO-4 bbr2d06 CASMO-4 bbr2d07 CASMO-4 bbr2d08 CASMO-4 bbr201 MCNP4a bbr202 MCNP4a bbr203 MCNP4a bbr210 MCNP4a bbr212 MCNP4a bbr216 MCNP4a Report: HI-982094 Computer studsvik PC PC PC PC PC PC PC PC PC PC PC PC DEC500au DEC500au DEC500au DEC500au DEC500au DEC500au DEC500au DEC500au DEC500au DEC500au DEC500au DEC500au DEC500au DEC500au PC PC PC PC PC PC Calclist-r2.xls Descnpt10n PWR REGION 1 CALCULATIONS 5.0 % Wl 7xl 7 for tolerances and temperature in Region 1 rack 5.0 wt% W17x17 -reference 5.0 wt% Wl 7xl 7 -eccentric positioning 5.0% W 17xl 7 - drop accident with 1 inch of exposed fuel 5.0% W 17x17 - drop accident with 2 inches of exposed fuel 5.0% W 17xl 7 - drop accident with 3 inches of exposed fuel 5.0% W 17xl 7 - drop accident with 4 inches of exposed fuel 5.0% W 17xl 7 - drop accident with 5 inches of exposed fuel 5.0 wt% Wl 7xl 7 - indepent check calculation for reference condition 5.0 wt% Wl 7xl 7 -5.5xl cell model with 1.38 in N-S rack-to-rack gap 5.0 wt% W17x17 -5.5xl cell model with l.651in N-S rack-to-rack gap 5.0 wt% W17x17 -5.5xl cell model with l.651in N-S and 1.337 in E-W rack-to-rack gap 5.0 wt% W17x17 -5.5xl cell model with l.38in N-S and 1.38 in E-W rack-to rack gap PWR REGION 2 CALCULATIONS 5.0% W 17xl 7 for tolerances & temperature at zero burnup 5.0% W 17xl 7 for tolerances & temperature at burnups of 35, 40, and 45 GWD/MTU 5.0% W 17xl 7 tolerance for higher fuel density at burnups of 35, 40, and 45 GWD/MTU 5.0% W 17x17 tolerance for lower fuel density at burnups of 35, 40, and 45 GWD/MTU 5.0% W 17xl7 tolerance for higher fuel enrichment at burnups of 35, 40, and 45 GWD/MTU 5.0% W 17xl 7 tolerance for lower fuel enrichment at burnups of 35, 40, and 45 GWD/MTU 5.0% W 17xl 7 in core depletion and restart in Region 2 rack 4.5% W 17xl 7 in core depletion and restart in Region 2 rack 4.0% W 17xl 7 in core depletion and restart in Region 2 rack 3.5% W 17xl 7 in core depletion and restart in Region 2 rack 3.0% W 17xl 7 in core depletion and restart in Region 2 rack 2.5 W l 7xl 7 in core depletion and restart in Region 2 rack 2.0% W 17xl 7 in core depletion and restart in Region 2 rack

[ ] W(axial blankets) W 17xl 7 in core depletion and restart in Region 2 rack 5.0% W l 7xl 7 burned to 40 GWD/MTU in Region 2 rack - reference case 5.0% W 17xl 7 burned to 40 GWD/MTU in Region 2 rack - eccentric fuel positioning case 5.0% W 17x17 burned to 40 GWD/MTU in Region 2 rack-infinite in z 5.0% W 17x17 burned to 30 GWD/MTU in Region 2 rack 5.0% W 17x17 assembly without axial blankets burned to 30 GWD/MTU with axial burnup distribution in Region 2 rack 5.0% W 17x17 assembly without axial blankets burned to 40 GWD/MTU with axial burnup distribution in Region 2 rack Holtec International Proprietary Information t? z, Page l -

Calclist-r2.xls Input ID Code Computer Description bbr221 MCNP4a PC 5.0% W l 7xl 7 assembly with axial blankets burned to 30 GWD/MTU with axial burnup distribution in Region 2 rack bbr225 MCNP4a PC 5.0% W 17xl7 assembly with axial blankets burned to 40 GWD/MTU with axial burnup distribution in Region 2 rack bbr230 MCNP4a PC 5x5 array of W l 7xl 7 with equiv enr for 5.0wt% burned to 40 GWD/MTU -

reference bbr231 MCNP4a PC 5x5 array of W 17xl 7 with equiv enr for 5.0wt% burned to 40 GWD/MTU -

with misplaced fresh 5.0 wt% assembly bbr232 MCNP4a PC 5x5 array of W 17xl 7 with equiv enr for 5.0wt% burned to 40 GWD/MTU -

with misplaced fresh 5.0 wt% assembly and 200 ppm soluble boron bbr233 MCNP4a PC 5x5 array of W 17xl 7 with equiv enr for 5.0wt% burned to 40 GWD/MTU -

with misplaced fresh 5.0 wt% assembly and 400 ppm soluble boron bbr234 MCNP4a PC 5x5 array ofW 17x17 with equiv enrfor 5.0wt% burned to 40 GWD/MTU -

with misplaced fresh 5.0 wt% assembly and 800 ppm soluble boron bbr240 MCNP4a PC 5.0% W 17xl 7 - drop accident with 1 inch of exposed fuel

. bbr241 MCNP4a PC 5.0% W 17x17 - drop accident with 2 inches of exposed fuel bbr242 MCNP4a PC 5.0% W 17xl 7 - drop accident with 3 inches of exposed fuel bbr243 MCNP4a PC 5.0% W 17xl 7 - drop accident with 4 inches of exposed fuel bbr244 MCNP4a PC 5.0% W 17xl 7 - drop accident with 5 inches of exposed fuel bbr250 MCNP4a PC mislocated fresh fuel assembly accident - reference bbr251 MCNP4a PC mislocated fresh fuel assembly accident - with mislocated assembly bbr252 MCNP4a PC mislocated fresh fuel assembly accident - with mislocated assembly and 200 ppm soluble boron bbr253 MCNP4a PC mislocated fresh fuel assembly accident - with mislocated assembly and 800 ppm soluble boron bbr254 MCNP4a PC mislocated fresh fuel assembly accident - with mislocated assembly and 400 ppm soluble boron kbbr2a KENO5a PC 5.0% W 17xl 7 burned to 40 GWD/MTU in Region 2 rack - independent verification calculation for the reference case Report: ffi-982094 Holtec International Proprietary Information E-3 Page __ _

Project No. 2216 Report No. HI-982094 Page S1-1 Holtec International Proprietary Information Supplement 1 Evaluation of Fuel Assemblies with Missing Fuel Rods for Storage in Region 2 (total number of pages 6)

Proprietary Information This appendix contains miscellaneous details of Constellation Energy and Holtec proprietary information, hence it is considered proprietary in its entirely.

Attachment A HOLTEC APPROVED COMPUTER PROGRAM LIST (total number of pages 2)

Project No. 2216 Report No. HI-982094R3 Page Al Heltee lfltematieaal Prearietltf'i lnfemtatiea

HOLTEC APPROVED COMPUTER PROGRAM LIST 1 REV.235 October 31, 2012 APPROVED IN CERTIFIED REMARKS: See OPERATING APPROVED Indicate PROGRAM USNRCPART VERSION USERS FOR "A" CODE report indicated SYSTEM&

COMPUTERS:

Computer (Category) 50 & 71/72 SER:

(Executable)

EXPERT for special VERSION (Docket#) 2 CODES limitations (Service pack4)

Listed by ID ID(s) used SP A, BDB, KB, 4-2.05.14 HF, SVF, TH, BK, SPA HI-2104750 Windows XP (3) 1006 1006 DMM, VIM, ES, PS SP A, BDB, KB, 5M-1.06.00 HF, SVF, TH,BK, SPA HI-2104750 Windows XP (2) 1008, 1013 CASMO(A)

DOC 50-271 DMM, VIM, ES, PS DOC 71-9336 SPA, BDB, KB, 5-2.00.00 HF, SVF, TH, BK, SPA HI-2104750 Windows 7 (0, 1) 1051 DMM, VIM, ES, PS SPA, BDB, KB, Windows 7 (0,1) 1051 5-2.02.00 HF, SVF, TH, BK, SPA HI-2104750 DMM, VIM, ES, PS Windows XP (2) 1008 Note 1: All codes on this list have been validated under Holtec's QA program.

Note 2: Programs with docket numbers listed have been identified by name in the SER by the USNRC. This column is for information only.

Note 3: Only computers identified on the ACPL may use SolidWorks for computing weights and center of gravity values on safety significant work. However, reports do not need to indicate which computers were used for the drawings referenced.

Note 4: A zero indicated as the service pack is equivalent to an operating system having no service pack.

Note 5: Only computers identified on the ACPL may use AutoCAD for computing volumes, moments of inertia and centroid values on safety significant work. However, reports do not need to indicate which computers were used for the drawings referenced.

Note 6: Certified Users for "A" Codes that have a "B" next to their name are only qualified to use the code on evaluations ofITS-B or lower categories of equipment.

Note 7: Keno is the only Module of Scale ver. 5.1 considered a Category A code requiring certified users.

Project No. 2216 Report No. HI-982094R3 PageA2 Holtee lfltemlttioaal Proorietlffi' lnformatioa