View source for RS-22-064, License Amendment Request Regarding New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies with Proposed Change to Technical Specifications Section 4.3.1

{{Adams
| number = ML22159A310
| issue date = 06/08/2022
| title = License Amendment Request Regarding New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies with Proposed Change to Technical Specifications Section 4.3.1
| author name = Simpson P
| author affiliation = Constellation Energy Company, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000237, 05000249
| license number = DPR-019, DPR-025
| contact person = 
| case reference number = RS-22-064
| package number = ML22159A309
| document type = Letter type:RS, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 85
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 RS-22-064                                                                                  10 CFR 50.90 June 8, 2022 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237, and 50-249

==Subject:==
License Amendment Request Regarding New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies with Proposed Change to Technical Specifications Section 4.3.1 In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) requests an amendment to Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS), Units 2 and 3, respectively. Specifically, CEG is adopting a new criticality safety analysis (CSA) methodology for performing the criticality safety evaluation for legacy fuel types in addition to the GNF3 reload fuel in the spent fuel pool (SFP). Use of the new SFP CSA methodology requires a change to the DNPS Technical Specifications (TS) 4.3.1, "Criticality."
CEG is also proposing a change to the new fuel vault (NFV) CSA to utilize the GESTAR II methodology for validating the NFV criticality safety for GNF3 fuel in the General Electric (GE) designed NFV racks.
A separate amendment request will be submitted to support the transition to GNF3 fuel at DNPS. While the revised SFP CSA and the altered NFV CSA basis support the planned transition to GNF3 fuel, neither the new analysis or the altered analysis basis is required to support the NRC review and approval of the separate fuel transition amendment request planned for submittal in late summer 2022. provides the evaluation of proposed changes. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 contains the non-proprietary version of NEDO-33938, "Dresden Nuclear Power Station Fuel Storage Criticality Safety Analysis," Revision 0, dated April 2022. Attachment 4 contains the Criticality Analysis Checklist from NEI 12-16, "Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants," Revision 4.
Attachment 6 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachment 6, this document is decontrolled.

June 8, 2022 U.S. Nuclear Regulatory Commission Page 2  contains proprietary information to be withheld from public disclosure in accordance with 10 CFR 2.390, as documented by the signed affidavit in Attachment 5. The affidavit sets forth the basis on which GNF's information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4),
"Public inspections, exemptions, requests for withholding." Accordingly, it is respectfully requested that the information, which is proprietary to GNF be withheld from public disclosure.
A non-proprietary version of the report is provided in Attachment 3.
The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using the criteria in 10 CFR 50.92(c), and it has been determined that the proposed changes involve no significant hazards consideration.
The proposed changes have been reviewed and approved by the Plant Operations Review Committee at DNPS in accordance with the requirements of the CEG Quality Assurance Program.
CEG requests approval of the proposed changes by June 8, 2023. Once approved, the amendment shall be implemented within 60 days.
In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"
paragraph (b), CEG is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Officials.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mrs. Linda M. Palutsis at (630) 657-2821.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 8th day of June 2022.
Respectfully, Patrick R. Simpson Sr. Manager Licensing Constellation Energy Generation, LLC Attachment 6 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachment 6, this document is decontrolled.

June 8, 2022 U.S. Nuclear Regulatory Commission Page 3 Attachments:
: 1. Evaluation of Proposed Changes
: 2. Markup of Proposed Technical Specification Pages
: 3. NEDO-33938, "Dresden Nuclear Power Station Fuel Storage Criticality Safety Analysis,"
Revision 0, dated April 2022 (Non-Proprietary Version)
: 4. NEI 12-16 Appendix C, Criticality Analysis Checklist
: 5. Global Nuclear Fuels - Americas, LLC 10 CFR 2.390 Affidavit
: 6. NEDC-33938P, "Dresden Nuclear Power Station Fuel Storage Criticality Safety Analysis," Revision 0, dated April 2022 (Proprietary Version) cc:    U.S. NRC Region III, Regional Administrator U.S. NRC Senior Resident Inspector, Dresden Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety Attachment 6 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachment 6, this document is decontrolled.

ATTACHMENT 1 Evaluation of Proposed Changes

==Subject:==
License Amendment Request Regarding New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies with Proposed Change to Technical Specifications Section 4.3.1 TABLE OF CONTENTS 1.0    

==SUMMARY==
DESCRIPTION 2.0    DETAILED DESCRIPTION 2.1   Spent Fuel Pool Criticality Safety Analysis 2.2   New Fuel Vault Criticality Safety Analysis 2.3   Proposed Changes to Technical Specifications Section 4.3.1

==3.0    TECHNICAL EVALUATION==

3.1   Spent Fuel Pool Criticality Safety Analysis 3.1.1   Overview of System Design and Operation 3.1.2   Criticality Evaluation 3.1.3   Accident Conditions 3.2   New Fuel Vault Criticality Safety Analysis

==4.0    REGULATORY EVALUATION==

4.1   Applicable Regulatory Requirements/Criteria 4.2   Precedent 4.3   No Significant Hazards Consideration 4.4   Conclusions

==5.0    ENVIRONMENTAL CONSIDERATION==

==6.0    REFERENCES==

Page 1 of 13

ATTACHMENT 1 Evaluation of Proposed Changes 1.0        

==SUMMARY==
DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) requests an amendment to Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS), Units 2 and 3, respectively. Specifically, CEG is adopting a new criticality safety analysis (CSA) methodology (Reference 6.1) for performing the criticality safety evaluation for legacy fuel types in addition to the GNF3 reload fuel in the spent fuel pool (SFP).
Use of the new SFP CSA methodology requires a change to the DNPS Technical Specifications (TS) 4.3.1, "Criticality." CEG is also proposing a change to the new fuel vault (NFV) CSA to utilize the GESTAR II methodology (Reference 6.2) for validating the NFV criticality safety for GNF3 fuel in the General Electric (GE) designed NFV racks.
2.0        DETAILED DESCRIPTION 2.1      Spent Fuel Pool Criticality Safety Analysis CEG will be transitioning from the ATRIUM 10XM fuel design to the GNF3 fuel design at DNPS beginning in the fall of 2023. The previous SFP CSA (see Reference 6.5) was prepared by Holtec International Inc. (Holtec). The CSA for the DNPS SFPs is being re-baselined by GNF to:
* Simplify the validation of GNF3 fuel designs against the CSA criteria. The new analysis will move DNPS away from the need to validate the in-rack kinf value for each new lattice design to now validating the standard cold core geometry (SCCG) kinf value against the defined limit. The SCCG kinf value is generated for every lattice in each assembly design as part of the standard calculation set.
* Provide a higher level of consistency among the BWR criticality safety analyses of record (AOR) methods utilized across the fleet. This also includes the methods utilized to verify new GNF3 fuel designs against the criticality safety AOR limitations as listed in the TS.
The reason for this license amendment is the re-baselined SFP CSAs change from Holtec methodology to GNF methodology. The proposed methodology change requires NRC approval prior to using the CSA in support of storage of fuel in the DNPS Unit 2 and Unit 3 SFPs. The DNPS Unit 2 and Unit 3 SFP racks are designed to accommodate BWR fuel. Both pools SFP racks credit Boral panels as the neutron absorbing material. The revised analysis shows that the effective neutron multiplication factor (keff) in the SFP racks fully loaded with fuel of the highest anticipated reactivity, at a temperature corresponding to the highest reactivity, does not exceed the 10 CFR 50.68(b) regulatory limit of 0.95 at a 95 percent probability, 95 percent confidence level. Reactivity effects of abnormal and accident conditions are also evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit.
Page 2 of 13

ATTACHMENT 1 Evaluation of Proposed Changes The SFP analysis is performed consistent with 10 CFR 50.68 requirements and industry guidance, including Nuclear Energy Institute (NEI) Report 12-16, Revision 4, "Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants,"
(Reference 6.4). Guidance pertaining to soluble boron in the SFP is not applicable because DNPS is a BWR plant and has no soluble boron in the SFP. The calculations are performed using GNFs method of analyzing SCCG kinf values and in-rack kinf values and validating the linear correlation between these parameters across a wide range of kinf values. The method then demonstrates that maintaining all fuel below the chosen SCCG kinf upper limit results in an in rack keff value no greater than 0.95 after accounting for biases and uncertainties (i.e., kmax (95/95)  0.95). The revised CSA covers all legacy fuel in storage in either the DNPS Unit 2 or Unit 3 SFP and the new GNF3 product line. A copy of the NEI 12-16 Criticality Analysis Checklist is included as Attachment 4 to identify the areas of the analysis that conform or do not conform to the guidance in NEI 12-16. Additional information is provided for any deviation from NEI 12-16 in the Attachment 4 checklist.
The change in the SFP CSA AOR requires a change to TS Section 4.3.1, "Criticality." The specifics of the TS change are provided in Section 2.3.
2.2      New Fuel Vault Criticality Safety Analysis The DNPS NFV racks are GE designed low density racks with a center-to-center spacing within a given row of 6.625 inches and an interrack spacing of 11.06 inches. The NFV rack CSA coverage for the new GNF3 fuel will be the GESTAR II (Reference 6.2) analysis for GE designed low density NFV racks, upon approval of this proposed license amendment. The previous NFV CSA will no longer be applicable to DNPS upon implementation of this license amendment because the only fuel to be delivered to the site for core reloads will be GNF3.
No TS change is needed for implementation of the GESTAR II NFV CSA methodology. The SCCG limit of kinf  1.31 is the GESTAR II basis NFV CSA limit for DNPS storage of fresh GNF3 fuel.
Section 3.2 provides additional details on the applicability of the GESTAR II NFV CSA methodology.
2.3      Proposed Changes to Technical Specifications Section 4.3.1 The DNPS, Units 2 and 3, TS requirements related to spent fuel storage are contained in TS Section 4.3, "Fuel Storage." TS 4.3.1, "Criticality," identifies requirements pertaining to the design of the SFP storage racks. Specifically, TS 4.3.1.1.a requires keff to be  0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1.2 of the Updated Final Safety Analysis Report (UFSAR). TS 4.3.1.1.b requires a nominal 6.30-inch center to center distance between fuel assemblies placed in the SFP storage racks in both pools. Neither of these sections require update because of the proposed change in CSA methodology.
The governing kinf limit structure for acceptable SFP fuel storage in TS 4.3.1.1.c is replaced with a new condition in accordance with the new CSA basis as shown below.
Page 3 of 13

ATTACHMENT 1 Evaluation of Proposed Changes Current TS 4.3.1.1.c                             Proposed TS 4.3.1.1.c The combination of U-235 enrichment and             Fuel assemblies having a maximum kinf of gadolinia loading shall be limited to ensure        1.33 in the normal reactor core configuration fuel assemblies have a maximum k-infinity of        at cold conditions.
0.8895 as determined at 39.2&#xba;F in the normal spent fuel pool in-rack configuration.
A mark-up of the proposed TS change is provided in Attachment 2. There are no TS Bases associated with Chapter 4, "Design Features." The DNPS UFSAR will be updated in accordance with 10 CFR 50.71(e) as part of implementation of the approved amendment. A summary of the proposed changes is provided below.
* An Issue Report was written on UFSAR Section 9.1.1.2, "Facilities Description," to correct an error discovered in the stated dimensions of the NFV. The UFSAR section is being updated to the correct dimensions, as shown on construction drawings of the DNPS NFV. The NFV storage dimensions have not been changed since construction of DNPS.
* Section 9.1.1.3, "Safety Evaluation," will be modified to reflect the NFV requirements of 10 CFR 50.68(b)(2) in this section and update cross-reference to ATRIUM 10XM fuel with GNF3 fuel.
* Section 9.1.2.3.1, "High-Density Fuel Racks," will be modified to reflect the characteristics of the new SFP CSA covering all fuel types.
* Section 9.1.5, "References," will be updated for consistency with other changes made in Section 9.1.

==3.0        TECHNICAL EVALUATION==

3.1      Spent Fuel Pool Criticality Safety Analysis 3.1.1      Overview of System Design and Operation The DNPS UFSAR Section 9.1.2, "Spent Fuel Storage," documents the combined units SFP safety design bases as follows:
A. There will be no release of contamination or exposure of personnel to radiation in excess of 10 CFR 20 limits; B. The storage space in each of the Unit 2 and Unit 3 spent fuel pools is designed for a maximum of 3537 irradiated fuel assemblies; C. It is possible, at any time, to perform limited work on irradiated components; D. Space is provided for used control rods, flow channels, and other reactor components; E. The spent fuel pool is designed to withstand earthquake loadings of a Class I structure; and F. The spent fuel assembly racks are designed to ensure subcriticality in the storage pool.
A maximum keff of 0.95 is maintained with the racks fully loaded with fuel of the highest Page 4 of 13

ATTACHMENT 1 Evaluation of Proposed Changes anticipated reactivity and flooded with unborated water at a temperature corresponding to the highest reactivity.
To achieve the safety design bases DNPS has two separate SFPs which provide for storage of unirradiated and irradiated fuel in a safe manner.
The DNPS SFPs are identical in the types of SFP racks and neutron absorbing materials used.
Each spent fuel pool contains 33 high-density spent fuel storage racks which provide storage for 3537 fuel assemblies. There are 18 racks arranged in a 9x11 array and 15 racks arranged in a 9x13 array. The racks are constructed to form tubes of adequate size for fuel storage. The tubes are welded together along their length with angles or clips to provide the inter-tube connection. The center-to-center distance between assemblies stored in tubes is 6.30 in. x 6.30 in.
The fuel storage tube is constructed of stainless-steel-bearing Boral neutron absorbing material.
Boral is a sandwich-type plate that has outer surfaces of Type 1100 aluminum and a core of boron carbide (B4C) uniformly dispersed in a matrix of Type 1100 aluminum. These plates are enclosed by inner and outer tubes made of Type 304 stainless steel designed to permit spent fuel pool water to enter and exit the Boral area. The inner and outer tubes maintain the Boral plate structural integrity during vibratory events. The plates are not required to carry load. The individual neutron absorbing tubes are connected in a checkerboard pattern forming the rack assembly.
Each rack consists of a base assembly with legs and with plates along the edges and across the midpoint. A fuel support plate fabricated from 3/16-inch plate is provided in each storage position to hold one fuel assembly. The support plate is elevated 12.37 inches above the spent fuel pool floor and is welded to the lower end of the tube. Cooling water flows through holes and/or slots in the sides of the support plates into the storage tubes to cool the stored fuel.
Along the side of the rack, a filler plate assembly is welded between the absorber tube assemblies to enclose the space between neutron absorbing tubes. The racks are designed to prevent application of excessive vertical forces from the fuel handling system.
The specific Boral panels used at DNPS have a minimum certified 10B areal density of 0.020 g/cm2.
The spent fuel storage racks are designed to maintain the stored spent fuel in a spatial geometry that precludes the possibility of criticality. The spent fuel storage racks maintain this subcritical geometry when subjected to maximum earthquake conditions, dropped fuel assembly accident conditions, and any uplift forces generated by the fuel handling equipment.
3.1.2      Criticality Evaluation In accordance with 10 CFR 50.68, the CSA for the DNPS Unit 2 SFP and Unit 3 SFP has been revised to support the purposes discussed in Section 2.1. The analysis, provided as , demonstrates that the maximum keff (i.e., kmax(95/95)) is less than the 10 CFR 50.68 limit of 0.95 for normal and credible abnormal operation with tolerances and computational uncertainties considered. All necessary requirements as outlined in Page 5 of 13

ATTACHMENT 1 Evaluation of Proposed Changes NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," Section 9.1.1, Revision 3 dated March 2007, have been met. NEI 12-16, Revision 4 (Reference 6.4) was used as guidance for this analysis.
The revised CSA covers all legacy fuel in storage in either the DNPS Unit 2 or Unit 3 SFP and the new GNF3 product line. The description of the GNF3 product line is provided in Section 4.1 of Attachment 6, while the description of legacy fuel is provided in Appendix B of Attachment 6.
The calculations are performed using GNFs peak in-core kinf methodology. The peak in-core kinf criterion method relies on a well-characterized relationship between the infinite lattice kinf (in-core) for a given fuel design and a specific fuel storage rack kinf (in-rack) containing that fuel.
This methodology was shown to be appropriate for use at DNPS by validating that there exists a well-characterized, linear relationship between the infinite lattice kinf (in-core) and fuel storage rack kinf (in-rack). Appropriate application was also ensured by using a design basis lattice with conservative values of rack efficiency and in-core kinf for all criticality analyses.
Appendix B of Attachment 6 shows that this method produces an in-core kinf which correlates to an in-rack kinf for GNF3 fuel that bounds the legacy fuel. This is in line with the requirements in 10 CFR 50.68(b) and NEI 12-16, Revision 4 (Reference 6.4). The CSA uses the minimum certified 10B areal density of 0.020 g/cm2 in the Boral containing spent fuel racks at DNPS.
The peak reactivity of the fuel in the DNPS SFP storage racks was calculated using the computer codes TGBLA06 and MCNP-05P. In this evaluation, in-core kinf values and exposure dependent, pin-by-pin isotopic specifications were generated using TGBLA06, the NRC approved GE-Hitachi Nuclear Energy Americas, LLC (GEH)/GNF BWR lattice physics code.
The fuel storage criticality calculations were then performed using MCNP-05P, the GEH/GNF proprietary version of the Los Alamos National Laboratory Monte Carlo neutron transport code, MCNP5, using the TGBLA06 nuclide inventory as input. TGBLA06 uses ENDF/B-V cross-section data to perform coarse-mesh, broad-group, diffusion theory calculations. MCNP-05P uses ENDF/B-VII.0 pointwise (i.e., continuous) cross section data, and all reactions in the cross-section evaluation are considered. MCNP-05P has been validated and verified for spent fuel pool storage rack evaluations in accordance with the NUREG/CR-6698 guidance (included as part of Attachment 6). The method of analysis is discussed in greater detail in Section 3.0 of . Validation of the codes and libraries is described in Section 3.4 and Appendix A of Attachment 6.
The use of TGBLA06 for BWR core depletion calculations has been reviewed and accepted by the NRC as part of the approval of Reference 6.2. The use of TGBLA06 for GNF3 depletion calculations has been reviewed and approved by the NRC as part of the GESTAR II Amendment 49 approval (Reference 6.8). The NRC has also approved the MCNP-05P
/TGBLA06 code package for use in similar fuel pool criticality analyses for other licensees.
Reference 6.6 documents an example of one NRC approved use of this code package.
3.1.3      Accident Conditions The spent fuel rack configuration was analyzed for credible accident scenarios. The scenarios considered are presented in the bulleted list that follows and are discussed in Section 5.5.3 of Page 6 of 13

ATTACHMENT 1 Evaluation of Proposed Changes . Note that the "no Boral on rack periphery" case is conservatively treated as a normal condition bias in Section 5.5.2. Note that the base cases in Section 5.5.2 and in Section 5.5.3 are not intended to be the same.
* SFP temperature exceeding the normal range (moderator temperature/density changes)
* Dropped and dropped + damaged fuel assemblies
* Rack movement (seismic)
* Mislocated fuel assembly (assemblies in the wrong location outside a storage rack)
The criticality analysis for the storage of BWR assemblies in the DNPS SFP racks with Boral panels has been performed. The results for the normal condition show that keff is  0.95 with the storage racks fully loaded with fuel of the highest anticipated reactivity, at a temperature corresponding to the highest reactivity. The results for all accident conditions also show that keff is  0.95 with the storage racks fully loaded with fuel of the highest anticipated reactivity, at a temperature corresponding to the highest reactivity.
Reactivity effects of abnormal and accident conditions have been evaluated and assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95 with a 95 percent probability at a 95 percent confidence level.
3.2       New Fuel Vault Criticality Safety Analysis The DNPS UFSAR Section 9.1.1, "New Fuel Storage," documents the applicable combined units NFV safety design basis. New fuel is stored in a manner which precludes inadvertent criticality. The NFV racks are designed to maintain the stored spent fuel in a spatial geometry that precludes the possibility of criticality. The DNPS NFV racks are GE designed low density racks with a center-to-center spacing within a given row of 6.625 inches and an interrack spacing of 11.06 inches. The NFV at DNPS is a reinforced concrete Class I structure that ensures the spacing is maintained even in a seismic event.
The NFV rack CSA coverage for the new GNF3 fuel will be the GESTAR II (Reference 6.2) analysis for GE designed low density NFV racks, upon approval of this proposed license amendment. The applicability of GESTAR II to the GNF3 fuel type is documented in the GNF3 GESTAR II validation report (Reference 6.3). GNF3 and the NFV methodology was previously reviewed by the NRC as part of Amendment 37 to GESTAR and was incorporated first in GESTAR in Revision 24. Amendment 37 (Reference 6.10) RAI-3, which is included in the US Supplement to GESTAR II (Reference 6.2), Page US.B-181, asked for additional details regarding GESTAR kinf calculations and methodology. The NRC staff reviewed the fresh and irradiated fuel storage criteria proposed and the methodology used in the calculations in Amendment 37 to GESTAR II and issued a safety evaluation (Reference 6.10) that determined that the criteria and methodology are acceptable.
GNF performed a CSA for GNF3 fuel stored in a NFV to ensure the applicability of GESTAR II to GNF3 fuel. The two models used in the CSA for the GNF3 NFV have rectangular dimensions corresponding to two generic NFV designs. The NFV criticality analysis modeled the actual rectangular dimensions and tolerances of both options to determine the restrictions outlined in Page 7 of 13

ATTACHMENT 1 Evaluation of Proposed Changes GESTAR II. The limiting model bounds the geometry of the DNPS NFV racks; therefore, the racks are covered under GESTAR II. The DNPS NFV interrack pitch is  10.5 inches (the criteria listed in GESTAR II) and thus the racks may be utilized to store new GNF3 fuel with in-core SCCG kinf  1.31 (Reference 6.2, Section 3.5).

==4.0         REGULATORY EVALUATION==

4.1      Applicable Regulatory Requirements/Criteria The regulations in 10 CFR 50.36, "Technical Specifications," contain the requirements for the contents of TS. As required by 10 CFR 50.36(c)(4), "Design features," the TS will include design features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in 10 CFR 50.36(c)(1), (2), and (3). TS 4.0, "Design Features," provides the DNPS requirements for site location, the reactor core, and fuel storage, meeting the intent of 10 CFR 50.36(c)(4). The governing kinf limit structure for acceptable SFP fuel storage in TS 4.3.1.1.c is replaced with a new condition that is consistent with new CSA basis.
10 CFR 50.68, "Criticality accident requirements," paragraph (b)(4) states that the keff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. Further, paragraphs (b)(2) and (b)(3) state the equivalent neutron multiplication factor limit for the NFV, including the impact that "optimum moderation" scenario might have. The requirements stated include that the keff of the fresh fuel in the fresh fuel storage racks loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. The regulation also states that for the optimum moderation case the keff must not exceed 0.98 at a 95 percent probability, 95 percent confidence level. The optimum moderation case is not applicable to the DNPS NFV as it is a moderation controlled area, as stated in DNPS UFSAR Section 9.1.1.3 (Reference 6.7). The moderation control area limits the amount of hydrogenous material on the Refuel Floor, which contains the NFV, at DNPS. Administrative controls, as generally defined in SIL 152 (Reference 6.9), have been incorporated in the DNPS fire protection plans and fuel handling procedures. DNPS utilizes these standard industry practices to comply with SIL 152 to preclude the optimum moderation condition. These practices have been found acceptable to forgo specifically analyzing the optimum moderation condition stipulated in 10 CFR 50.68(b)(3).
The DNPS SFP criticality analysis, provided as Attachment 6 to this submittal and the DNPS NFV racks meeting the GESTAR II storage requirements summarized in Section 3.5 of Reference 6.2, demonstrate that these requirements are met.
10 CFR 50.68(b)(7) states that the maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to 5.0 percent by weight. DNPS GNF3 fuel is below 5.0 percent by weight U-235 enrichment.
DNPS, Units 2 and 3, were not licensed to the 10 CFR 50, Appendix A, General Design Criteria (GDC). The DNPS, Units 2 and 3, UFSAR, Section 3.1, "Conformance with NRC General Design Criteria," provides an assessment against the 70 draft GDC published in 1967 and concluded that the plant specific requirements are sufficiently similar to the Appendix A GDC.
Page 8 of 13

ATTACHMENT 1 Evaluation of Proposed Changes The design basis of Unit 2 was later evaluated against the Appendix A GDC in UFSAR Section 3.1.2. The high degree of similarity between Units 2 and 3 indicates that Unit 3 also conforms to the GDC. Criterion 66, "Prevention of fuel storage criticality," states that criticality in the new and spent fuel storage shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations. The evaluation of DNPS's conformance with Criterion 66 is discussed in both Section 3.1.1.8.1, "Criterion 66 - Prevention of Fuel Storage Criticality" and Section 9.1, "Fuel Storage and Handling," of the DNPS UFSAR. The racks in which new and spent fuel assemblies are placed are designed and arranged to ensure subcriticality in the vault and storage pool. The DNPS criticality analysis demonstrates that, given the current spent fuel storage system design, keff will remain less than or equal to 0.95 for the legacy fuel types and the GNF3 fuel.
4.2      Precedent The NRC recently approved the use of the GNF CSA methodology to determine the acceptability of storing fresh and spent GNF3 fuel at River Bend Station. Amendment No. 201 and the corresponding safety evaluation were issued on December 31, 2019 (Reference 6.6).
The difference between the River Bend Station submittal and this submittal is the neutron absorbing material credited in the spent fuel pool. River Bend Station utilizes NETCO-SNAP-IN rack inserts, and DNPS credits Boral panels for reactivity control. However, the GNF methodology is independent of the neutron absorbing material in the spent fuel pool racks. This GNF methodology was shown to be appropriate for use at DNPS by validating that there exists a well-characterized, linear relationship between the infinite lattice kinf (in-core) and fuel storage rack kinf (in-rack). Appropriate application was also ensured by using a design basis lattice with conservative values of rack efficiency and in-core kinf for all criticality analyses. Therefore, this methodology is applicable to DNPS.
4.3      No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) requests an amendment to Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS), Units 2 and 3, respectively. Specifically, CEG is adopting a new criticality safety analysis (CSA) methodology for performing the criticality safety evaluation for legacy fuel types in addition to the GNF3 reload fuel in the spent fuel pool (SFP). Use of the new SFP CSA methodology requires a change to the DNPS Technical Specifications (TS) 4.3.1, "Criticality."
CEG is also proposing a change to the new fuel vault (NFV) CSA to utilize the GESTAR II methodology for validating the NFV criticality safety for GNF3 fuel in the General Electric (GE) designed NFV racks. This methodology change for the NFV does not require a change to the DNPS TS.
CEG has evaluated the proposed change against the criteria of 10 CFR 50.92(c) to determine if the proposed change results in any significant hazards. The following is the evaluation of each of the 10 CFR 50.92(c) criteria:
: 1.       Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Page 9 of 13

ATTACHMENT 1 Evaluation of Proposed Changes Response: No The proposed amendment involves use of new methodologies for performing the new fuel vault (NFV) criticality safety analysis (CSA) and the spent fuel pool (SFP) CSA for the DNPS Units 2 and 3 spent fuel pools (SFPs). Technical Specification 4.3.1.1.c requires revision to maintain consistency with the new methodology results. The proposed new CSA demonstrates adequate margin to criticality and therefore does not affect the consequences of any accident previously evaluated.
The impact of the CSA methodology change was evaluated on the following four previously evaluated events and accidents: fuel handling accident (FHA), fuel mispositioning event, seismic event, and loss of SFP cooling event.
The proposed amendment, covering only the change in CSA methodologies, does not change or modify the fuel handling processes, new and spent fuel storage racks, number of fuel assemblies that may be stored in the NFV or SFP, the assumed decay heat generation rate, or the SFP cooling and cleanup system. There is therefore no impact on the probability of an accident previously evaluated.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No Onsite storage of spent and fresh fuel assemblies in the DNPS Unit 2 SFP and Unit 3 SFP and onsite storage of fresh fuel assemblies in a shared NFV are normal activities for which DNPS has been designed and licensed. The proposed use of new methodologies for performing the DNPS NFV CSA and SFP CSA does not change or modify the fuel handling processes, new or spent fuel racks, number of fuel assemblies that may be stored in the new fuel vault or spent fuel pools, decay heat generation rate, or the SFP cooling and cleanup system.
The fuel handling accident (FHA), fuel mispositioning event, seismic event, and loss of SFP cooling event do not represent a new or different type of accident. The associated analysis results show that the storage racks remain sub-critical following a worst-case FHA, fuel mispositioning event, seismic event, and loss of SFP cooling event. Note that the missing rack Boral panel event was conservatively modeled as part of the evaluation of normal conditions instead of as a separate accident.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
: 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No Page 10 of 13

ATTACHMENT 1 Evaluation of Proposed Changes DNPS TS 4.3, "Fuel Storage," Specification 4.3.1.1.a requires the spent fuel storage racks to maintain the effective neutron multiplication factor, keff, less than or equal to 0.95 when fully flooded with unborated water, which includes an allowance for uncertainties. Therefore, for spent fuel pool criticality considerations, the required safety margin is 5 percent. 10 CFR 50.68(b)(2) also requires a keff of less than or equal to 0.95 in the NFV (the optimum moderation, 10 CFR 50.68(b)(3), case does not apply to DNPS due to countermeasures taken to prevent water fog entry into the NFV). Thus, the NFV also has a required safety margin of 5 percent.
The proposed change ensures, as verified by the associated NFV and SFP criticality analyses, that keff continues to be less than or equal to 0.95, thus preserving the required safety margin of 5 percent. In addition, using the in-core kinf limit ensures that the SFP criticality analysis remains bounding for the fuel assemblies that are allowed to be stored in the SFP storage racks. The NFV and SFP criticality analyses provide adequate protection to ensure public health and safety in that it determines the reactivity limit for the fuel assemblies that are allowed to be stored in the NFV and SFP storage racks.
The proposed use of a new methodology for performing the DNPS SFP CSA does not affect spent fuel heat generation or the spent fuel cooling systems.
In addition, the radiological consequences of a dropped fuel assembly remain unchanged as the anticipated fuel damage due to a fuel handling accident is unaffected by using a new methodology to perform the CSA. The proposed change also does not increase the capacity of either the Unit 2 or Unit 3 spent fuel pools or the shared NFV beyond the current capacity.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, CEG concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
4.4      Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

==5.0        ENVIRONMENTAL CONSIDERATION==

The proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant Page 11 of 13

ATTACHMENT 1 Evaluation of Proposed Changes increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

==6.0      REFERENCES==

6.1     NEDC-33938P, "Dresden Nuclear Power Station Fuel Storage Criticality Safety Analysis," Revision 0, dated April 2022 (Attachment 3 (non-proprietary) and Attachment 6 (proprietary version) of this submittal) 6.2     GE Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR II, Main)," Revision 31, dated November 2020 (ADAMS Accession Number ML20330A197/ML20330A198 (proprietary version) and ML20330A199 (non-proprietary version))
6.3     Letter from B. Moore (Global Nuclear Fuel, Americas, LLC) to U. S. NRC, "Revised General Electric Standard Application for Reactor Fuel (GESTAR II) Compliance Reports for the GNF2 and GNF3 Fuel Product Lines," dated August 28, 2020 (ADAMS Accession Number ML20244A104) 6.4     NEI 12-16, "Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants," Revision 4, dated September 2019 (ADAMS Accession Number ML19269E069) 6.5     Letter from E. Brown (U. S. NRC) to B. Hanson (Exelon Generation Company, LLC),
        "Dresden Nuclear Power Station, Units 2 and 3 - Issuance of Amendments to Amend Renewed Facility Operating License Nos. DPR-19 and DPR-25 to Support Use of a New Nuclear Criticality Safety Analysis Methodology (CAC Nos. MF5734 and MF5735)," dated April 29, 2016 (ADAMS Accession Number ML15343A126) 6.6     Letter from M. O'Banion (U. S. NRC) to Entergy Operations, Inc., "River Bend Station, Unit 1 - Issuance of Amendment No. 201 RE: Change to the Neutron Absorbing Material Credited in Spent Fuel Pool for Criticality Control (EPID L-2018-LLA-0298),"
dated December 31, 2019 (ADAMS Accession Number ML19357A009) 6.7     Dresden Units 2 and 3 - Updated Final Safety Analysis Report (UFSAR), Revision 14, dated June 2021 6.8     Letter from D. Morey (U. S. NRC) to M. P. Catts (GE-Hitachi Nuclear Energy Americas, LLC), "Final Safety Evaluation for NEDC-33173P Supplement 5 - Applicability of GE Methods to Expanded Operating Domains - Supplement for GNF3 Fuel (EPID: L-2017-TOP-0033)," dated March 21, 2019 (ADAMS Accession Number ML19064A229 (proprietary version) and ML19074A054 (non-proprietary version))
Page 12 of 13

ATTACHMENT 1 Evaluation of Proposed Changes 6.9  GE SIL No. 152, "Criticality Margins for Storage of New Fuel," dated March 31, 1976 (Attachment A of ML20198C524) 6.10 Letter from K. Hsueh (U.S. NRC) to J. G. Head (GE-Hitachi Nuclear Energy Americas, LLC), "Final Safety Evaluation for Amendment 37 to Global Nuclear Fuel - Americas Topical Report NEDE-24011-P-A-US General Electric Standard Application for Reactor Fuel and the US Supplement (CAC NO. MF0743)," dated March 2017. (ADAMS Accession Number ML17066A291/ML17069A311)
Page 13 of 13

ATTACHMENT 2 Markup of Proposed Technical Specification Pages Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 REVISED TECHNICAL SPECIFICATIONS PAGES 4.0-2

Design Features 4.0 4.0   DESIGN FEATURES    (continued) 4.3   Fuel Storage 4.3.1    Criticality 4.3.1.1  The spent fuel storage racks are designed and shall be maintained with:
: a. keff  0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1.2 of the UFSAR; and Fuel assemblies having a maximum kinf of 1.33       b. A nominal 6.30 inch center to center distance in the normal reactor             between fuel assemblies placed in the storage core configuration at             racks.
cold conditions.
: c. The combination of U-235 enrichment and gadolinia loading shall be limited to ensure fuel assemblies have a maximum k-infinity of 0.8895 as determined at 39.2&#xba;F in the normal spent fuel pool in-rack configuration.
4.3.2    Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 589 ft 2.5 inches.
4.3.3    Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 3537 fuel assemblies.
Dresden 2 and 3                          4.0-2               Amendment No. 249/242

ATTACHMENT 3 NEDO-33938, "Dresden Nuclear Power Station Fuel Storage Criticality Safety Analysis,"
Revision 0, dated April 2022 (Non-Proprietary Version)

Global Nuclear Fuel NEDO-33938 Revision 0 April 2022 Non - Proprietary Information Dresden Nuclear Power Station Fuel Storage Criticality Safety Analysis Copyright 2022 Global Nuclear Fuel, All Rights Reserved

NEDO-33938 Revision 0 Non-Proprietary Information INFORMATION NOTICE This is a non-proprietary version of the document NEDC-33938P, Revision 0, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here ((           )).
IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document are furnished for the purpose of supporting Dresden Nuclear Power Station evaluation of spent fuel pool criticality. The use of this information by anyone other than Dresden Nuclear Power Station, or for any purpose other than that for which it is furnished by GNF is not authorized; and with respect to any unauthorized use, GNF makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.
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NEDO-33938 Revision 0 Non-Proprietary Information TABLE OF CONTENTS

==1.0 INTRODUCTION==
 ............................................................................................................... 1 2.0 REQUIREMENTS .............................................................................................................. 1 3.0 METHOD OF ANALYSIS ................................................................................................. 1 3.1  Cross-Sections ......................................................................................................................... 2 3.2  Geometry Treatment ................................................................................................................ 2 3.3  Convergence Checks ............................................................................................................... 3 3.4  Validation and Computational Basis ....................................................................................... 3 3.5  In-Core k Methodology .......................................................................................................... 6 3.6  Definitions ............................................................................................................................... 7 3.7  Assumptions and Conservatisms ............................................................................................. 8 4.0 FUEL DESIGN BASIS ........................................................................................................ 9 4.1  GNF3 Fuel Description............................................................................................................ 9 4.2  Fuel Model Description ......................................................................................................... 13 5.0 CRITICALITY ANALYSIS OF SPENT FUEL STORAGE RACKS ......................... 14 5.1  Description of Spent Fuel Storage Racks .............................................................................. 14 5.2  Spent Fuel Storage Rack Models ........................................................................................... 17 5.3  Design Basis Lattice Selection .............................................................................................. 19 5.4  Normal Configuration Analysis ............................................................................................. 21 5.4.1 Analytical Models ...................................................................................................... 21 5.4.2 Results ........................................................................................................................ 22 5.5  Bias Cases .............................................................................................................................. 22 5.5.1 Depletion Bias Cases ................................................................................................. 22 5.5.2 Normal Bias Cases ..................................................................................................... 23 5.5.3 Abnormal/Accident Bias Cases ................................................................................. 24 5.5.4 Results ........................................................................................................................ 26 5.6  Uncertainties .......................................................................................................................... 27 5.6.1 Analytical Models ...................................................................................................... 27 5.6.2 Results ........................................................................................................................ 28 5.7  Maximum Reactivity ............................................................................................................. 30 6.0 INTERFACES BETWEEN STORAGE POOLS ........................................................... 31

==7.0 CONCLUSION==
S ................................................................................................................ 31

==8.0 REFERENCES==
 .................................................................................................................. 32 APPENDIX A MCNP-05P CODE VALIDATION ................................................................ 33 A.1 Trend Analysis ......................................................................................................................... 38 A.2 Bias and Bias Uncertainty Calculation - Single Sided Tolerance Limit.................................. 42 APPENDIX B LEGACY FUEL STORAGE JUSTIFICATION .......................................... 45 B.1 Legacy Non-GNF Fuel Justification ........................................................................................ 45 B.2 Legacy GNF Fuel Justification ................................................................................................. 46 iv

NEDO-33938 Revision 0 Non-Proprietary Information LIST OF TABLES Table 1 - Summary kmax(95/95) Result ......................................................................................... 1 Table 2 - Summary of the Critical Benchmark Experiments ........................................................ 4 Table 3 - Area of Applicability Covered by Code Validation ....................................................... 5 Table 4 - Lattice Dimensions....................................................................................................... 11 Table 5 - Cell Dimensions ........................................................................................................... 11 Table 6 - Channel Dimensions .................................................................................................... 12 Table 7 - Fuel Stack Density as a Function of Gadolinia Concentration .................................... 13 Table 8 - Storage Rack Model Dimensions ................................................................................. 19 Table 9 - GNF3 Fuel Parameter Ranges Studied in Spent Fuel Rack ......................................... 20 Table 10 - In-Rack k Results - Normal Configurations ............................................................ 22 Table 11 - Rack Periphery Study Results .................................................................................... 24 Table 12 - Results for Misplaced Bundles................................................................................... 25 Table 13 - Spent Fuel Storage Rack Bias Summary ................................................................... 27 Table 14 - Spent Fuel Storage Rack Uncertainty k Results ...................................................... 29 Table 15 - Spent Fuel Storage Rack Results Summary ............................................................... 30 Table 16 - MCNP-05P Results for the Benchmark Calculations ................................................ 33 Table 17 - Trending Parameters .................................................................................................. 38 Table 18 - Trending Results Summary ........................................................................................ 41 Table 19 - Bias and Bias Uncertainty for MCNP-05P with ENDF/B-VII .................................. 43 Table 20 - Recommended Bias and Bias Uncertainty in Criticality Analyses for MCNP-05P with ENDF/B-VII ................................................................................................................................. 44 Table 21 - Summary Non-GNF Fuel k Compared to GNF3 Design Basis ............................... 45 Table 22 - Limiting Cold As-Designed Eigenvalue of all GNF Bundles .................................... 46 v

NEDO-33938 Revision 0 Non-Proprietary Information LIST OF FIGURES Figure 1 - GNF3 Lattice Configuration ....................................................................................... 10 Figure 2 - Channel 1/8 Cross-Sections ........................................................................................ 12 Figure 3 - GNF3 MID Lattice in MCNP-05P.............................................................................. 14 Figure 4 - Spent Fuel Storage Pool Arrangement ........................................................................ 15 Figure 5 - Stainless Steel Tube with Boral Core ......................................................................... 16 Figure 6 - High Density Spent Fuel Rack .................................................................................... 17 Figure 7 - Storage Rack Model Schematic .................................................................................. 18 Figure 8 - Zoomed Storage Rack Model Schematic.................................................................... 18 Figure 9 - Spent Fuel In-Rack versus In-Core Eigenvalues ........................................................ 21 Figure 10 - Finite Misplaced Bundle Model Example ................................................................ 25 Figure 11 - Scatterplot of knorm versus EALF ........................................................................... 39 Figure 12 - Scatterplot of knorm versus wt.% 235U ....................................................................... 39 Figure 13 - Scatterplot of knorm versus wt.% 239Pu ...................................................................... 40 Figure 14 - Scatterplot of knorm versus H/X ................................................................................ 40 Figure 15 - Normality Test of knorm Results ................................................................................ 42 vi

NEDO-33938 Revision 0 Non-Proprietary Information The fuel storage criticality calculations are then performed using MCNP-05P, the GEH/GNF proprietary version of MCNP5 (Reference 4). MCNP-05P is a Monte Carlo program for solving the linear neutron transport equation for a fixed source or an eigenvalue problem. The code implements the Monte Carlo process for neutron, photon, electron, or coupled transport involving all these particles, and computes the eigenvalue for neutron-multiplying systems. For the present application, only neutron transport was considered.
3.1        Cross-Sections TGBLA06 uses ENDF/B-V cross-section data to perform coarse-mesh, broad-group, diffusion theory calculations. It includes thermal neutron scattering with hydrogen using an S(,) light water thermal scattering kernel.
MCNP-05P uses pointwise (i.e., continuous) cross-section data, and all reactions in a given cross-section evaluation (e.g., ENDF/B-VII.0) are considered. For the present work, thermal neutron scattering with hydrogen was described using S(,) light water thermal scattering kernel. The cross-section tables include all details of the ENDF representations for neutron data. The code requires that all the cross-sections be given on a single union energy grid suitable for linear interpolation; however, the cross-section energy grid varies from isotope to isotope. The libraries include very little data thinning and utilize resonance integral reconstruction error tolerances of 0.001%.
3.2        Geometry Treatment TGBLA06 is a 2D lattice design computer program for Boiling Water Reactor (BWR) fuel bundle analysis. It assumes that a lattice is uniform and infinite along the axial direction and that the lattice geometry and material are reflecting with respect to the lattice boundary along the transverse directions.
MCNP-05P implements a robust geometry representation that can correctly model complex components in three dimensions. An arbitrary three-dimensional configuration is treated as geometric cells bounded by first and second-degree surfaces and some special fourth-degree elliptical tori. The cells are described in a cartesian coordinate system and are defined by the intersections, unions and complements of the regions bounded by the surfaces. Surfaces are defined by supplying coefficients to the analytic surface equations or, for certain types of surfaces, known points on the surfaces. Rather than combining several pre-defined geometrical bodies in a combinatorial geometry scheme, MCNP-05P has the flexibility of defining geometrical shapes from all the first and second-degree surfaces of analytical geometry and elliptical tori and then combining them with Boolean operators. The code performs extensive checking for geometry errors and provides a plotting feature for examining the geometry and material assignments.
2

NEDO-33938 Revision 0 Non-Proprietary Information 3.3       Convergence Checks The use of TGBLA06 as a depletion code in this criticality analysis is consistent with its use for BWR fuel design and its associated users manual. Convergence checks are encoded in the standard error routines and the absence of error messages was confirmed in all code output.
In this analysis, the following criticality code parameters were specified. At a minimum, all MCNP-05P cases were run with 20,000 neutrons per generation, 200 cycles skipped, and 500 total cycles run. Some cases were run for more cycles skipped in order to meet all the convergence checks. For this analysis, the following MCNP-05P convergence checks were reviewed and confirmed passed for each case:
* Sampling of all cells that contain fissionable material
* Matching of first and second half eigenvalue
* Fission source entropy check 3.4       Validation and Computational Basis MCNP-05P has been compared to ((                  )) critical experiments for validation purposes using ENDF/B-VII.0 nuclear cross-section data. The experiments cover a number of moderator-to-fuel ratios and poison materials that represent material and geometric properties similar to that of BWR fuel lattices both in and out of fuel racks. The critical experiments to which MCNP-05P has been compared are provided in Table 2. All are either low-enriched Uranium Dioxide (UO2) or Mixed Uranium-Plutonium Oxide (MOX) pin lattice in water experiments. The area of applicability (AOA) considered covered by this validation is listed in Table 3, along with the parameters which characterize the spent fuel rack system for comparison. The critical experiment modeling results, along with the calculation of the associated bias and bias uncertainty terms at the 95/95 confidence level using NUREG-6698 (Reference 5) guidance are provided in Appendix A. The study concluded that the appropriate bias to apply to systems covered by this AOA is ((           )), and the appropriate uncertainty of that bias is ((           )).
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* TGBLA06 eigenvalue uncertainty An additional uncertainty is also added to the fuel rack study related to eigenvalue calculations performed using TGBLA06. A bias of ((              ))and a 95/95 bias uncertainty of ((
                                                                                   )) This uncertainty is applied to the spent fuel racks kmax(95/95) value to cover uncertainty in the assignment of in-core k values to fuel lattices.
3.5        In-Core k Methodology The design of the fuel storage racks provides for a subcritical multiplication factor for both normal and credible abnormal storage conditions. In all cases, the storage rack eigenvalue must be  0.95.
To demonstrate compliance with this limit, the in-core k method is utilized.
The peak in-core k criterion method relies on a well-characterized relationship between infinite lattice k (in-core) for a given fuel design and a specific fuel storage rack k (in-rack) containing that fuel. The use of an infinite lattice k criterion for demonstrating compliance to fuel storage criticality criteria has been used for all General Electric-supplied storage racks and is currently used for re-rack designs at several plants. This report demonstrates that the methodology is also appropriate for use at the DNPS by presenting the following:
* A well-characterized, linear relationship between infinite lattice k (in-core) and fuel storage rack k (in-rack)
* The use of a design basis lattice with a conservative rack efficiency and in-core k for all criticality analyses The analysis performed to calculate the lattice k to confirm compliance with the above criterion uses the Nuclear Regulatory Commission (NRC)-approved lattice physics methods encoded into the TGBLA06 engineering computer program. One of the outputs of the TGBLA06 solution is the lattice k of a specific nuclear design for a given set of input state parameters (e.g., void fraction, control state, fuel temperature).
Compliance of fuel with specified k limits will be confirmed for each new lattice as part of the bundle design process. Documentation that this has been met will be contained in the fuel design information report, which defines the maximum lattice k for each assembly nuclear design. The process for validating that specific assembly designs are acceptable for storage in the DNPS fuel storage racks is provided below.
: 1.       Identify the unique lattices in each assembly design.
: 2.       Deplete the lattices in TGBLA06 using the following conditions:
: a. Assembly aligned according to DNPS-specific lattice spacing and zero leakage
: b. ((
c.
6

NEDO-33938 Revision 0 Non-Proprietary Information d.
e.
i.
ii.
f.
g.
3.
a.
b.
c.
d.
e.
: f.                                        ))
: 4.      Ensure that the k values obtained from Step 3 for each lattice are less than or equal to the k limit of 1.33.
3.6      Definitions Fuel Assembly - is a complete fuel unit consisting of a basic fuel rod structure that may include large central water rods. Several shorter rods may be included in the assembly. These are called part-length rods. A fuel assembly includes the fuel channel.
Gadolinia - The compound Gd2O3. The gadolinium content in integral burnable absorber fuel rods is usually expressed in weight percentage gadolinia.
Lattice - An axial zone of a fuel assembly within which the nuclear characteristics of the individual rods are unchanged.
Base Lattice (BAS)- An axial zone of a fuel assembly located in the bottom half of the bundle within which all possible fuel rod locations for a given fuel design are occupied.
Mid Lattice (MID) ((
                                                                           ))
Vanished Lattice (VAN) - An axial zone of a fuel assembly typically in the upper half of the bundle within which a number of possible fuel rod locations are unoccupied.
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NEDO-33938 Revision 0 Non-Proprietary Information Rack Efficiency - The ratio of a particular lattice statepoint in-rack eigenvalue (k) to its associated lattice nominal in-core eigenvalue (k). This value allows for a straightforward comparison of a racks criticality response to varying lattice designs within a particular fuel product line. A lower rack efficiency implies increased reactivity suppression capability relative to an alternate design with a higher rack efficiency.
Design Basis Lattice - The lattice geometry, exposure history, and corresponding fuel isotopics for a fuel product line that result in the highest rack efficiency in a sensitivity study of reasonable fuel parameters at the desired in-core reactivity. This lattice is used for all normal, abnormal, and tolerance evaluations in the fuel rack analysis.
3.7       Assumptions and Conservatisms The fuel storage rack criticality calculations are performed with the following assumptions to ensure the true system reactivity is always less than the calculated reactivity:
: 1. ((
                                               ))
: 2. ((
                                                                                                   ))
: 3. Design basis lattices with in-core k values greater than the proposed 1.33 in-core k limit are used for all criticality analyses.
: 4. ((
                                                                                                       ))
Sensitivity studies of the storage system reactivity to these depletion parameters are presented in Section 5.5. ((
         ))
: 5. For conservatism, only positive reactivity differences from nominal conditions determined from depletion sensitivity and abnormal configuration analyses are added as biases to the final storage rack kmax(95/95).
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: 6. Neutron absorption in spacer grids, concrete, activated corrosion and wear products (CRUD), and axial blankets is ignored to limit parasitic losses in non-fuel materials.
: 7. TGBLA06 defined lumped fission products and Xe-135 are both conservatively ignored for MCNP-05P in-rack k calculations.
: 8. ((
                                               ))
: 9. Only 10B is modeled in the rack poison panel. Each panel is assumed to contain the minimum areal density, 0.020 g10B/cm2. All other material is ignored. Ignoring the other materials conservatively limits neutron absorption in the panel.
4.0       FUEL DESIGN BASIS This rack criticality analysis covers the GNF3 fuel product line and all legacy fuel stored at DNPS.
Justification for the storage of all legacy fuel at DNPS is provided in Appendix B. The description of GNF3 fuel is in Section 4.1. The GNF3 fuel bundle is used to determine the design basis bundle in Section 5.3.
All fuel is UO2 with some fuel rods containing gadolinia, Gd2O3.
This criticality analysis covers reconstituted fuel where a rod containing fuel is replaced with another fueled or non-fueled rod. Fuel where there are missing fuel rod locations that are not part of the normal fuel product line designs is the non-standard fissile configurations assessed in Section 5.5.3.
This criticality analysis also covers the storage of non-fuel items such as channels in spent fuel rack locations because this analysis covers peak reactivity fuel in every rack cell location.
4.1       GNF3 Fuel Description The GNF3 fuel lattice configuration is a 10x10 fuel rod array ((
                                                                                       )), as shown in Figure 1 with corresponding dimensions in Table 4 and Table 5. Figure 1 also demonstrates the part-length rod locations. Fuel channel dimensions are provided in Figure 2 and Table 6. Pellet stack density is in Table 7. ((
                                                                                               ))
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((
                                         ))
Figure 1 - GNF3 Lattice Configuration 10

NEDO-33938 Revision 0 Non-Proprietary Information Table 7 - Fuel Stack Density as a Function of Gadolinia Concentration Gadolinia Concentration         ((
(wt. fraction)
Pellet Density
                                                                                                          ))
(g/cc) 4.2       Fuel Model Description The fuel models considered include 2D geometric modeling of all fuel material, cladding, water rods, and channels. In the depletion model, appropriate depletion time steps are used consistent with depletion timesteps used in BWR core design analyses. ((
                               )) Pin specific isotopic modeling as a function of exposure is performed based on the lattice physics code TGBLA06. To obtain the isotopic composition of the fuel pins, each lattice design considered is burned at reactor operating ((
                                                           )) and depleted through to a final exposure of
((
                      )) The isotopics utilized exclude Xe-135 and TGBLA06 defined lumped fission products ((
            )) An example of a GNF3 MID lattice model in MCNP-05P is depicted in Figure 3.
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((
                                                                                           ))
Figure 3 - GNF3 MID Lattice in MCNP-05P The fuel loadings considered for each lattice span a range of exposures, average enrichments, number of gadolinia rods, gadolinia concentrations, and void histories considered to be reasonably representative of any DNPS fuel design. The lattice type and exposure history that results in the worst-case rack efficiency for an in-core k greater than the proposed limit is then used to define the design basis lattice. This lattice is assumed to be stored in every location in the rack being analyzed. Details on the determination of the design basis lattice using the process outlined above are presented in Section 5.3.
5.0      CRITICALITY ANALYSIS OF SPENT FUEL STORAGE RACKS 5.1      Description of Spent Fuel Storage Racks Each spent fuel pool contains 33 high-density spent fuel storage racks which provide storage for 3537 fuel assemblies. There are 18 racks arranged in a 9x11 array and 15 racks arranged in a 9x13 array. See Figure 4 for the relative location of the racks in each spent fuel pool. The racks are constructed to form tubes of adequate size for fuel storage. The tubes are welded together along their length with angles or clips to provide the inter-tube connection. The center-to-center distance between assemblies stored in tubes is 6.30 in. x 6.30 in.
The fuel storage tube is constructed of stainless-steel-bearing Boral neutron absorbing material.
Boral is a sandwich-type plate (see Figure 5) that has outer surfaces of Type 1100 aluminum and a core of boron carbide (B4C) uniformly dispersed in a matrix of Type 1100 aluminum. These plates are enclosed by inner and outer tubes made of Type 304 stainless steel designed to permit spent fuel pool water to enter and exit the Boral area. The inner and outer tubes maintain the Boral plate 14

NEDO-33938 Revision 0 Non-Proprietary Information structural integrity during vibratory events. The plates are not required to carry load. The individual neutron absorbing tubes are connected in a checkerboard pattern forming the rack assembly.
The rack assembly is shown in Figure 6. Each rack consists of a base assembly with legs and with plates along the edges and across the midpoint. A fuel support plate fabricated from 3/16-inch plate is provided in each storage position to hold one fuel assembly. The support plate is elevated approximately 12.37 inches above the spent fuel pool floor and is welded to the lower end of the tube. Cooling water flows through holes and/or slots in the sides of the support plates into the storage tubes to cool the stored fuel. Along the side of the rack, a filler plate assembly is welded between the absorber tube assemblies to enclose the space between neutron absorbing tubes. The racks are designed to prevent application of excessive vertical forces from the fuel handling system.
Figure 4 - Spent Fuel Storage Pool Arrangement 15

NEDO-33938 Revision 0 Non-Proprietary Information Figure 5 - Stainless Steel Tube with Boral Core 16

NEDO-33938 Revision 0 Non-Proprietary Information Figure 6 - High Density Spent Fuel Rack 5.2      Spent Fuel Storage Rack Models This analysis covers a bounding storage configuration of maximum reactivity fuel in every storage location for DNPS rack design.
A 2D infinite 9x11 storage rack array with periodic boundary conditions is modeled to conservatively represent the nominal spent fuel pool configuration. Cells with Boral panels and cells without Boral panels alternate in a checkerboard pattern. The model is based on adjacent racks having edge cells without Boral panels lined-up together. Images of the model are provided in 17

NEDO-33938 Revision 0 Non-Proprietary Information Figures 7 and 8. Dimensions and tolerances are presented in Table 8. This array is used in the design basis bundle selection process in Section 5.3.
((
                                                               ))
((
                                                                                    ))
Figure 7 - Storage Rack Model Schematic
((
                                                                                ))
Figure 8 - Zoomed Storage Rack Model Schematic 18

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((
                                                                                              ))
Figure 9 - Spent Fuel In-Rack versus In-Core Eigenvalues 5.4      Normal Configuration Analysis 5.4.1  Analytical Models The most reactive normal configuration was determined by studying the reactivity effect of the following credible normal scenarios:
* Storage of non-channeled assemblies
* Eccentric loadings o When neutron absorber panels with an areal density above 0.01 g 10B/cm2 are present on all four sides of the fuel assembly, a centrally located positioning of the fuel assembly in the storage cell is the most reactive configuration. Therefore, no eccentric loading cases were performed consistent with NEI 12-16.
    *  ((
o o                                            ))
* Pool moderator temperature variation 21

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                                                      ))
The following potential reactivity effect of changes that occur during depletion are considered:
: a. Fuel rod changes (clad creep, fuel densification/swelling)
Clad Creep - ((
                                                                                                 ))
Fuel Pellet Densification - ((
                                                   ))
: b. Material dependent grid growth
((
                                                         ))
5.5.2   Normal Bias Cases The following bias cases are included for normal conditions. As noted in Table 10, (from Section 5.4.2) cases with positive reactivity increases from the nominal case are included in roll-up of kBias and are therefore included in Table 13.
* No Boral on rack periphery There may be assemblies on the sides that are not surrounded by neutron absorbing Boral panels. ((
                                                                                        )) Results are provided in Table 11. The reactivity increase from this study is included in the final kBias term.
23

NEDO-33938 Revision 0 Non-Proprietary Information
* Rack sliding due to seismic event which causes water gap between racks to close The racks modeled in this analysis are infinite in extent with no inter-module water gaps.
This essentially assumes all racks are close-fitting and bounds possible reactivity effects of rack sliding.
* Loss of spent fuel pool cooling Normal sensitivity analysis results demonstrate that system reactivity decreases as moderator density decreases and pool temperature increases; therefore, reactivity effects of loss of spent fuel pool cooling are bounded by the nominal reactivity results.
* Storage of non-standard fissile configurations Non-standard fissile configurations of Optima2 fuel bundle DSB084 are stored along with standard fuel in the high density storage rack. Specifically, container SUB_B_TETBA001 contains DSB084 sub-bundle B stored in the Unit 3 spent fuel pool and original DSB084 fuel bundle contains sub-bundles A, C, and D stored in the Unit 3 spent fuel pool. Those scenarios are bounded by the limiting legacy non-GNF fuel which is addressed in Appendix B.
5.5.4    Results The results of the bias studies are provided in Table 13. The k term in the table represents the difference between the system reactivity with the specified bias case and kNormal. kB6 is the MCNP-05P bias from Section 3.4. The total contribution from these independent conditions to the kmax(95/95) of the spent fuel rack is calculated using Equation 1. In this equation, each k value must be both positive and the largest for its respective term to be considered.
n k Bias =  k Bi i =1                        (1) 26

NEDO-33938 Revision 0 Non-Proprietary Information Rod cladding thickness decreased by ((` ` ` ` ` ` ` )) and rod cladding outer diameter decrease by
((` ` ` ` ` ` ))
Channel thickness increase by ((` ` ` ` ` ` ))
Channel thickness decrease by ((` ` ` ` ` ` ))
Fuel pellet outer diameter increase by ((` ` ` ` ` ` ` ` ` ))
Fuel pellet outer diameter decrease by ((` ` ` ` ` ` ` ` ` ))
Fuel rod pin pitch increase by ((` ` ` ` ` ` ` ` ))
Fuel rod pin pitch decrease by ((` ` ` ` ` ` ` ` ))
Rack wall thickness increase by 0.030 inches (inner wall), 0.043 inches (outer wall)
Rack wall thickness decrease by 0.030 inches (inner wall), 0.060 inches (outer wall)
Rack pitch increase by 0.060 inches Rack pitch decrease by 0.060 inches The models developed for these studies were all based on the normal configuration presented in Section 5.4.
5.6.2     Results The results of the tolerance studies and uncertainties are provided in Table 14. The values are summed using Equation 2 which is adopted from NEI 12-16 (Reference 3).
The kTi term in the table represents the difference between the system reactivity with the specified tolerance perturbation and kNormal. In Equation 2, a kTi value must be both positive and the largest for its respective term to be considered.
The kUi terms in the table represent the uncertainty contributions to kmax(95/95) of the spent fuel rack and from the problem and code specific uncertainties which are combined with the tolerance contributions (kTi) using Equation 2.
2                2 1          1 (2) 28

NEDO-33938 Revision 0 Non-Proprietary Information 6.0      INTERFACES BETWEEN STORAGE POOLS The Unit 2 and Unit 3 storage pools are neutronically decoupled because the pools are not connected.

==7.0      CONCLUSION==
S The DNPS spent fuel racks have been analyzed for the storage of GNF3 fuel using the MCNP-05P Monte Carlo neutron transport program and the k criterion methodology. A maximum SCCG, uncontrolled peak in-core eigenvalue (k) of 1.33 as defined by TGBLA06 is specified as the rack design limit for GNF3 fuel in the spent fuel racks. The analyses resulted in a storage rack maximum k-effective (kmax(95/95)) less than the 10 CFR 50.68 limit of 0.95 for normal and credible abnormal operation with tolerances and computational uncertainties taken into account. Justification for the continued storage of all legacy DNPS fuel types is in Appendix B.
31

NEDO-33938 Revision 0 Non-Proprietary Information

==8.0      REFERENCES==
: 1) MFN-035-99, S. Richards (NRC) to G. Watford (GE), Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, "GESTAR II" - Implementing Improved GE Steady State Methods (TAC No. MA6481), November 10, 1999.
: 2) NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, US NRC, Revision 3, March 2007. (NRC ADAMS Accession Number ML070570006).
: 3) NEI 12-16 Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants, Revision 4, September 2019. (NRC ADAMS Accession Number ML19269E069).
: 4) LA-UR-03-1987, MCNP - A General Monte Carlo N-Particle Transport Code, Version 5, April 2003.
: 5) NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, US NRC, January 2001. (NRC ADAMS Accession Number ML050250061).
: 6) J.R. Taylor, An Introduction to Error Analysis, page 268-271, 2nd Edition, University Science Books, 1997.
32

NEDO-33938 Revision 0 Non-Proprietary Information
((
                                                        ))
Figure 11 - Scatterplot of knorm versus EALF
((
                                                     ))
Figure 12 - Scatterplot of knorm versus wt.% 235U 39

NEDO-33938 Revision 0 Non-Proprietary Information
((
                                                      ))
Figure 13 - Scatterplot of knorm versus wt.% 239Pu
((
                                                      ))
Figure 14 - Scatterplot of knorm versus H/X 40

NEDO-33938 Revision 0 Non-Proprietary Information A.2 Bias and Bias Uncertainty Calculation - Single Sided Tolerance Limit As no trends are apparent in the critical experiment results, a weighted single-sided tolerance limit methodology is utilized to establish the bias and bias uncertainty for this AOA and code package combination. Use of this method requires the critical experiment results to have a normal statistical distribution. This was verified using the Anderson-Darling normality test. A graphical image of the results for this normality test, including the p-value for the distribution, is provided in Figure 15.
Because the reported p-value is greater than 0.05, it is confirmed that the data fits a normal distribution, and the single sided tolerance limit methodology is confirmed to be applicable.
((
                                                                                              ))
Figure 15 - Normality Test of knorm Results When using this method, the weighted bias and bias uncertainty are calculated using the following equations:
                           =   1                     (A-5)
                            =  *    (A-6) 42

NEDO-33938 Revision 0 Non-Proprietary Information n
knorm i i =1      t2 k norm  =          n 1
i =1  t 2
(A-7)
SP = s2 +                  2 (A-8) n 2 =      n 1
2 i =1      t                               (A-9) 2 1 n 1 2 (k norm i  k norm )
2     n  1  i =1  t s =
1 n 1 n i =1  t2 (A-10)
Where:
k norm = Average weighted knorm S P = Pooled standard deviation s 2 = Variance about the mean 2 = Average total variance U = one-sided tolerance factor for n data points at (95/95 confidence/probability level) n = number of data points ((                  ))
Table 19 summarizes the results of these calculations.
Table 19 - Bias and Bias Uncertainty for MCNP-05P with ENDF/B-VII Bias (weighted)                                  ((
Bias Uncertainty (95/95 level)
Variance About the Mean Average Total Variance Pooled Standard Deviation (1)
One-Sided Tolerance Factor                                 ))
43

NEDO-33938 Revision 0 Non-Proprietary Information Using the average weighted bias and pooled standard deviation; the upper one-sided 95/95-tolerance limit (bias uncertainty) was calculated for use in criticality calculations, in accordance with NUREG-6698 guidance. As seen in Figure 15, ((
                           )) As shown in Table 19, the MCNP-05P bias uncertainty (95/95) ((
                                                                            )) Table 20 summarizes the recommended bias and bias uncertainty to be used in criticality calculations.
Table 20 - Recommended Bias and Bias Uncertainty in Criticality Analyses for MCNP-05P with ENDF/B-VII Bias                                ((
Bias Uncertainty (95/95)                    ))
44

ATTACHMENT 4 NEI 12-16 Appendix C, Criticality Analysis Checklist

APPENDIX C: CRITICALITY ANALYSIS CHECKLIST The criticality analysis checklist is completed by the applicant prior to submittal to the NRC. It provides a useful guide to the applicant to ensure that all the applicable subject areas are addressed in the application, or to provide justification/identification of alternative approaches.
The checklist also assists the NRC reviewer in identifying areas of the analysis that conform or do not conform to the guidance in NEI 12-16. Subsequently, the NRC review can then be more efficiently focused on those areas that deviate from NEI 12-16 and the justification for those deviations.
Subject                          Included                Notes / Explanation 1.0 Introduction and Overview Purpose of submittal                                     YES        Section 1.0 of NEDC-33938P and Section 1.0 of LAR RS-22-064.
Changes requested                                        YES        Section 1.0 of NEDC-33938P and Sections 2.1, 2.2, and 2.3 of LAR RS-22-064.
Summary of physical changes                      YES        Section 1.0 of NEDC-33938P and Sections 2.1, 2.2, and 2.3 of LAR RS-22-064.
Summary of Tech Spec changes                     YES        Section 2.3 of LAR RS-22-064.
Attachment 2 and Attachment 3 of LAR RS-22-064.
Summary of analytical scope                      YES        Sections 1.0 and 3.0 of NEDC-33938P and Sections 2.1 and 2.2 of LAR RS-22-064.
2.0 Acceptance Criteria and Regulatory Guidance Summary of requirements and guidance                     YES        Section 2.0 of NEDC-33938P and Sections 2.1 and 2.2 of LAR RS-22-064.
Requirements documents referenced                YES        Section 2.0 of NEDC-33938P and Sections 2.1 and 2.2 of LAR RS-22-064.
Guidance documents referenced                    YES        Section 2.0 of NEDC-33938P and Sections 2.1 and 2.2 of LAR RS-22-064.
Acceptance criteria described                    YES        Section 2.0 of NEDC-33938P and Sections 2.1 and 2.2 of LAR RS-22-064.
3.0 Reactor and Fuel Design Description C-1

Subject                    Included           Notes / Explanation Describe reactor operating parameters             NO    Not applicable for this analysis. See Sections 3.7 and 5.5 of NEDC-33938P for depletion parameters and assumptions.
Describe all fuel in pool                        YES    Section 4.0 and Appendix B of NEDC-33938P.
Geometric dimensions (Nominal and         YES    Section 4.1 of NEDC-33938P.
Tolerances)
Schematic of guide tube patterns           NO    Not applicable for BWR fuel; therefore, not applicable to NEDC-33938P.
Material compositions                     YES    Section 4.0 of NEDC-33938P.
Describe future fuel to be covered               YES    Section 4.0 of NEDC-33938P.
Geometric dimensions (Nominal and         YES    Section 4.1 NEDC-33938P.
Tolerances)
Schematic of guide tube patterns           NO    Not applicable for BWR fuel; therefore, not applicable to NEDC-33938P.
Material compositions                     YES    Section 4.0 of NEDC-33938P.
Describe all fuel inserts                         NO    Not applicable for BWR fuel; Geometric Dimensions (Nominal and                therefore, not applicable to Tolerances)                                      NEDC-33938P.
Schematic (axial/cross-section)
Material compositions Describe non-standard fuel                       YES    Sections 4.0 and 5.5.3 of Geometric dimensions                             NEDC-33938P.
Describe non-fuel items in fuel cells            YES    Section 4.0 of NEDC-33938P.
Nominal and tolerance dimensions           NO    Not applicable; analysis NEDC-33938P covers peak reactivity in every spent fuel pool rack cell location.
4.0 Spent Fuel Pool/Storage Rack Description New fuel vault & Storage rack description        YES    The new fuel vault analysis will be Nominal and tolerance dimensions                 covered by the GESTAR II Schematic (axial/cross-section)                  methodology and is not addressed in Material compositions                            NEDC-33938P. See Section 2.2 of LAR RS-22-064 for details.
Spent fuel pool, Storage rack description        YES    Sections 5.1-5.2 of NEDC-33938P and Nominal and tolerance dimensions                 Section 3.1 of LAR RS-22-064.
Schematic (axial/cross-section)
Material compositions Other Reactivity Control Devices (Inserts)       YES    Sections 5.1-5.2 of NEDC-33938P and Nominal and tolerance dimensions                 Section 3.1 of LAR RS-22-064.
Schematic (axial/cross-section)
Material compositions C-2

Subject                     Included           Notes / Explanation 5.0 Overview of the Method of Analysis New fuel rack analysis description                YES    The new fuel vault analysis will be Storage geometries                                covered by the GESTAR II Bounding assembly design(s)                       methodology and is not addressed in Integral absorber credit                          NEDC-33938P. See Section 2.2 of Accident analysis                                 LAR RS-22-064 for details.
Spent fuel storage rack analysis description      YES    Sections 3.5-3.7 and 5.0 of NEDC-33938P.
Storage geometries                         YES    Sections 5.1-5.2 of NEDC-33938P and Section 3.1 of LAR RS-22-064.
Bounding assembly design(s)                YES    Section 5.3 of NEDC-33938P.
Soluble boron credit                        NO    Not applicable - No soluble boron credit in BWR criticality analysis Boron dilution analysis                   NEDC-33938P.
Burnup credit                               NO    No burnup credit in BWR peak reactivity analysis NEDC-33938P -
fuel is evaluated at peak reactivity Decay/Cooling time credit                   NO    No decay/cooling time credit in NEDC-33938P.
Integral absorber credit                   YES    Sections 5.1-5.2 of NEDC-33938P.
Other credit                                NO    No other credit in analysis NEDC-33938P.
Fixed neutron absorbers                    YES    Boral panels - Sections 5.1-5.2 of NEDC-33938P and Section 3.1 of LAR RS-22-064.
Aging management program            NO    Aging is not included in this analysis.
Boral coupon surveillance program is covered by a previously approved license amendment.
Accident analysis                          YES    Section 5.5.3 of NEDC-33938P.
Temperature increase               YES    Sections 5.4-5.5 of NEDC-33938P.
Assembly drop                      YES    Section 5.5.3 of NEDC-33938P.
Single assembly misload            YES    Section 5.5.3 of NEDC-33938P.
Multiple misload                    NO    The Dresden Unit 2 and Unit 3 spent fuel pools are uniform pools. There is no opportunity for multiple misload.
Boron dilution                      NO    Not applicable - No soluble boron credit in BWR criticality analysis NEDC-33938P.
Other                              YES    Sections 5.5.3 of NEDC-33938P.
Fuel out of rack analysis                  YES    Section 5.5.3 of NEDC-33938P Handling                                  considers the worst case abnormal Movement                                  positioning of a fuel assembly outside a Inspection                                storage rack.
C-3

Subject                     Included            Notes / Explanation 6.0 Computer Codes, Cross Sections and Validation Overview Code/Modules Used for Calculation of keff         YES    Section 3.0 of NEDC-33938P.
Cross section library                      YES    Section 3.1 of NEDC-33938P.
Description of nuclides used               YES    Section 4.2 of NEDC-33938P.
Convergence checks                         YES    Section 3.3 of NEDC-33938P.
Code/Module Used for Depletion                    YES    Section 3.0 of NEDC-33938P.
Calculation Cross section library                      YES    Section 3.1 of NEDC-33938P.
Description of nuclides used               YES    Section 4.2 of NEDC-33938P.
Convergence checks                         YES    Section 3.3 of NEDC-33938P.
Validation of Code and Library                    YES    Section 3.4 and Appendix A of NEDC-33938P.
Major Actinides and Structural             YES    Section 3.4 of NEDC-33938P.
Materials Minor Actinides and Fission Products       YES    Section 3.4 of NEDC-33938P.
Absorbers Credited                         YES    Section 3.4 of NEDC-33938P.
7.0 Criticality Safety Analysis of the New Fuel Rack Rack model                                        YES    The new fuel vault (NFV) rack Boundary conditions                               criticality safety analysis coverage for Source distribution                               the new GNF3 fuel will be the Geometry restrictions                                    GESTAR II analysis for the GE Limiting fuel design                                     designed low density NFV racks upon Fuel density                                      approval of this license amendment.
Burnable Poisons                                  The Dresden NFV racks are GE Fuel dimensions                                   designed low density racks with an Axial blankets                                    interrack spacing of ~11 inches, which Limiting rack model                                      is  10.5 inches (the criteria listed in Storage vault dimensions and materials            GESTAR II) and thus, the racks may Temperature                                       be utilized to store new GNF fuel with in-core SCCG kinf  1.31. See Section Multiple regions/configurations 2.2 of LAR RS-22-064 for details.
Flooded Low density moderator Eccentric fuel placement Tolerances Fuel geometry Fuel pin pitch Fuel pellet OD Fuel clad OD Fuel content Enrichment Density Integral absorber C-4

Subject                     Included           Notes / Explanation Rack geometry Rack pitch Cell wall thickness Storage vault dimensions/materials Code uncertainty Biases Temperature Code bias Moderator Conditions Fully flooded and optimum density moderator 8.0 Depletion Analysis for Spent Fuel Depletion Model Considerations                     YES   Sections 3.0, 3.3, 3.4, 3.7, and 4.2 of Time step verification                            NEDC-33938P.
Convergence verification                  (NO for Simplifications                           Soluble Soluble Boron not applicable to BWR Non-uniform enrichments                    Boron) analysis.
Post Depletion Nuclide Adjustment Cooling Time Depletion Parameters Burnable Absorbers Integral Absorbers Soluble Boron Fuel and Moderator Temperature Power Control rod insertion Atypical Cycle Operating History 9.0 Criticality Safety Analysis of Spent Fuel Pool Storage Racks Rack model                                         YES   Section 5.2 of NEDC-33938P.
Boundary conditions Source distribution Geometry restrictions Design Basis Fuel Description                      YES   Section 5.3 of NEDC-33938P.
Fuel density                                YES   Section 4.1 of NEDC-33938P.
Burnable Poisons                            YES   Sections 4.1, 4.2, and 5.3 of NEDC-33938P.
Fuel assembly inserts                       NO    No fuel assembly inserts in analysis NEDC-33938P and none in use at Dresden.
Fuel dimensions                             YES   Section 4.1 of NEDC-33938P.
Axial blankets                              NO    Section 3.7 of NEDC-33938P.
Configurations considered                   YES   Section 5.4 of NEDC-33938P.
C-5

Subject                   Included           Notes / Explanation Borated                           NO    Not applicable for BWR analysis NEDC-33938P.
Unborated                        YES    BWR analysis NEDC-33938P considers the spent fuel pool to be unborated.
Multiple rack designs             NO    NEDC-33938P utilizes only one spent fuel rack design because only one is present at Dresden. The NFV racks are covered by the GESTAR II analysis (see Section 2.2 of LAR RS-22-064).
Alternate storage geometry        NO    Not applicable for analysis NEDC-33938P.
Reactivity Control Devices                     YES    Sections 5.1-5.2 of NEDC-33938P.
Fuel Assembly Inserts                    NO    No fuel assembly inserts in analysis NEDC-33938P or in use at Dresden.
Storage Cell Inserts                    YES    Sections 5.1-5.2 of NEDC-33938P contain information on Boral panels (no rack inserts)
Storage Cell Blocking Devices            NO    No blocking devices in analysis NEDC-33938P or in use at Dresden.
Axial burnup shapes                             NO    Section 3.7 of NEDC-33938P.
Uniform/Distributed                     YES    Section 3.7 of NEDC-33938P.
Nodalization                             NO    Section 3.7 of NEDC-33938P.
Blankets modeled                         NO    Section 3.7 of NEDC-33938P.
Tolerances/Uncertainties                       YES    Sections 5.6 and 5.7 of NEDC-33938P.
Fuel geometry Fuel rod pin pitch Fuel pellet OD Cladding OD Axial fuel position               NO    Section 3.7 of NEDC-33938P.
Fuel content                            YES    Section 5.6 of NEDC-33938P.
Enrichment Density Assembly insert dimensions and           NO    No fuel assembly inserts in analysis materials                                      NEDC-33938P or in use at Dresden.
C-6

Subject                     Included           Notes / Explanation Rack geometry                              YES    Section 5.6 of NEDC-33938P.
Flux-trap size (width)              NO    Not applicable to non-flux-trap racks, such as those in the Dresden SFP.
Rack cell pitch                    YES    Section 5.6 of NEDC-33938P.
Rack wall thickness                YES    Section 5.6 of NEDC-33938P.
Neutron Absorber Dimensions        YES    Section 5.6 of NEDC-33938P.
Rack insert dimensions and materials        NO    No rack inserts in NEDC-33938P and none in use at Dresden station.
Code validation uncertainty                YES    Sections 3.4, 5.6, and Appendix A of NEDC-33938P.
Criticality case uncertainty               YES    Section 5.6 of NEDC-33938P.
Depletion Uncertainty                      YES    Sections 3.4, 5.8 Burnup Uncertainty                          NO    Not applicable for BWR peak reactivity analysis NEDC-33938P.
Biases                                            YES    Section 5.0 of NEDC-33938P.
Design Basis Fuel design                   YES    Section 5.3 of NEDC-33938P.
Code bias                                  YES    Sections 3.4, 5.5 of NEDC-33938P.
Temperature                                YES    Section 5.4 of NEDC-33938P.
Eccentric fuel placement                   YES    Sections 5.4-5.5 of NEDC-33938P.
Incore thimble depletion effect             NO    Not applicable for analysis NEDC-33938P NRC administrative margin                   NO    Not applicable for analysis NEDC-33938P Modeling simplifications                          YES    Sections 3.7, 4.2 of NEDC-33938P.
Identified and described 10.0 Interface Analysis Interface configurations analyzed                 YES    Sections 5.5, 6.0 of NEDC-33938P.
Between dissimilar racks                   YES    Section 6.0 of NEDC-33938P.
Between storage configurations within      YES    Section 5.5 of NEDC-33938P.
a rack Interface restrictions                             NO    Section 6.0 of NEDC-33938P.
11.0 Normal Conditions Fuel handling equipment                            NO    Not in the scope and does not impact results of criticality analysis NEDC-33938P.
Administrative controls                            NO    No new administrative controls included in NEDC-33938P or LAR RS-22-064.
Fuel inspection equipment or processes             NO    Not in the scope and does not impact results of criticality analysis NEDC-33938P.
Fuel reconstitution                               YES    Section 4.0 of NEDC-33938P.
C-7

Subject                   Included           Notes / Explanation 12.0 Accident Analysis Boron dilution                                  NO    Not applicable - No soluble boron Normal conditions                              credit in BWR criticality analysis Accident conditions                            NEDC-33938P.
Single assembly misload                        YES    Section 5.5.3 of NEDC-33938P.
Fuel assembly misplacement                     YES    Section 5.5.3 of NEDC-33938P.
Neutron Absorber Insert Misload                 NO    Not applicable as there are no neutron absorber inserts; however, no Boral on rack periphery is evaluated in Section 5.5.3 of NEDC-33938P.
Multiple fuel misloads                          NO    Uniform pool, single storage configuration, no opportunity for multiple misloads.
Dropped assembly                               YES    Section 5.5.3 of NEDC-33938P.
Temperature                                    YES    Sections 5.4 and 5.5 of NEDC-33938P.
Seismic event/other natural phenomena          YES    Section 5.5.3 of NEDC-33938P.
13.0 Analysis Results and Conclusions Summary of results                             YES    Section 7.0 of NEDC-33938P.
NO    Not applicable for BWR peak reactivity Burnup curve(s) analyses, including NEDC-33938P.
NO    Not applicable for BWR peak reactivity Intermediate Decay time treatment analyses, including NEDC-33938P.
NO    No new administrative controls New administrative controls                           included in NEDC-33938P or LAR RS-22-064.
Technical Specification markups                YES    Section 2.3 of LAR RS-22-064.
Attachment 2 of LAR RS-22-064.
14.0 References                                YES    Section 8.0 of NEDC-33938P.
Appendix A: Computer Code Validation:                 Appendix A of NEDC-33938P.
Code validation methodology and bases          YES    Appendix A of NEDC-33938P.
New Fuel Depleted Fuel MOX HTC Convergence Trends Bias and uncertainty Range of applicability                  YES    Described in Section 3.4 of NEDC-33938P.
Analysis of Area of Applicability       YES    Described in Section 3.4 of coverage                                       NEDC-33938P.
C-8

ATTACHMENT 5 Global Nuclear Fuels - Americas, LLC 10 CFR 2.390 Affidavit

Global Nuclear Fuel - Americas AFFIDAVIT I, Brian R. Moore, state as follows:
(1) I am General Manager, Core & Fuel Engineering, Global Nuclear Fuel - Americas, LLC (GNF-A), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.
(2) The information sought to be withheld is contained in GNF-A proprietary report, NEDC-33938P, Dresden Nuclear Power Station Fuel Storage Criticality Safety Analysis, Revision 0, April 2022. GNF-A proprietary information within the text and tables is identified by a dotted underline placed within double square brackets.
((This sentence is an example.{3})) Figures and large objects containing GNF-A proprietary information are identified with double square brackets before and after the object. In all cases, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.
(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GNF-A 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 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought 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, 975 F2d 871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F2d 1280 (DC Cir. 1983).
(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 GNF-A's competitors without license from GNF-A 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 aspects of past, present, or future GNF-A customer-funded development plans and programs, resulting in potential products to GNF-A;
: d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
NEDC-33938P Revision 0                                                  Affidavit Page 1 of 3

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) To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GNF-A, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GNF-A, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions 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, or subject to the terms under which it was licensed to GNF-A.
(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 delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GNF-A 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 identified in paragraph (2), above, is classified as proprietary because it contains details of GNF-As fuel design and licensing methodology. The development of this methodology, along with the testing, development and approval was achieved at a significant cost to GNF-A or its licensor.
The development of the fuel design and licensing methodology along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GNF-A asset.
(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF-A's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GNF-A's comprehensive BWR safety and 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. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.
NEDC-33938P Revision 0                                                Affidavit Page 2 of 3

The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GNF-A.
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.
GNF-A's competitive advantage will be lost if its competitors are able to use the results of the GNF-A 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 GNF-A 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 GNF-A of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on this 28th day of April 2022.
Brian R. Moore General Manager, Core & Fuel Engineering Global Nuclear Fuel - Americas, LLC 3901 Castle Hayne Road Wilmington, NC 28401 Brian.Moore@ge.com NEDC-33938P Revision 0                                                  Affidavit Page 3 of 3}}