Text

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
	ML21298A168
Person / Time
Site:	Quad Cities
Issue date:	10/25/2021
From:	Simpson P; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML21298A167	List: RS-21-065, 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;
References
	RS-21-065 NEDO-33932, Rev 1
	Download: ML21298A168 (89)
	v • d • e

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 RS-21-065 10 CFR 50.90 October 25, 2021 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

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

Reference:

Public Pre-submittal Meeting Between Exelon Generation Company, LLC and the U.S. Nuclear Regulatory Commission, "Proposed License Amendment Request Associated with the Transition to a New Fuel Type and Vendor at LaSalle Station, Units 1 & 2 and Quad Cities Nuclear Power Station, Units 1 & 2,"

May 27, 2021 (ADAMS Accession Nos. ML21133A167 & ML21141A010)

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, respectively. Specifically, EGC is utilizing 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 necessitates a change to the QCNPS Technical Specifications (TS) 4.3.1, "Criticality." EGC 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.

EGC participated in a pre-submittal meeting with the NRC (see Reference) regarding the planned transition from the Framatome ATRIUM 10XM fuel design to the Global Nuclear Fuels -

America, LLC (GNF) GNF3 fuel design at QCNPS. During this meeting, EGC's plans to submit this amendment request supporting both the use of a new CSA methodology for performing the criticality safety evaluation in the spent fuel pools and returning to the GNF GESTAR II coverage as the CSA basis in the NFV were discussed. A separate amendment request will be Attachment 7 contains Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachment 7, this document is decontrolled.

October 25, 2021 U.S. Nuclear Regulatory Commission Page 2 submitted to support the transition to GNF3 fuel at QCNPS. 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 2021.

The following attachments are included in support of this proposed license amendment: : Evaluation of Proposed Changes : Mark-up of QCNPS, Units 1 and 2 Technical Specifications Pages : NEDO-33932, "Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis," Revision 1, dated October 2021 (Non-Proprietary Version) : NEI 12-16 Criticality Analysis Checklist : GNF 10 CFR 2.390 Affidavit : Curtiss-Wright Corporation (Curtiss-Wright) 10 CFR 2.390 Affidavit : NEDC-33932P, "Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis," Revision 1, dated October 2021 (Proprietary Version) contains information proprietary to GNF and Curtiss-Wright. As a result, this document is supported by signed affidavits from the owners of the information, which are included as Attachments 5 and 6, respectively. Each affidavit sets forth the basis on which the corporations information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, it is respectfully requested that the information, which is proprietary to GNF and Curtiss-Wright be withheld from public disclosure. A redacted 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 by the QCNPS Plant Operations Review Committee in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed changes by October 25, 2022. Once approved, the amendment will be implemented within 60 days. This implementation period will provide adequate time for the affected station documents to be revised using the appropriate change control mechanisms.

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

paragraph (b), a copy of this letter and its attachments, is being provided to the designated State Officials.

October 25, 2021 U.S. Nuclear Regulatory Commission Page 3 There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Ms. Rebecca L. Steinman at (630) 657-2831.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 25th day of October 2021.

Respectfully, Patrick R. Simpson Sr. Manager Licensing Exelon Generation Company, LLC Attachments:

1. Evaluation of Proposed Changes

2. Mark-up of QCNPS, Units 1 and 2 Technical Specifications Pages

3. NEDO-33932, "Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis,"

Revision 1, dated October 2021 (Non-Proprietary Version)

4. NEI 12-16 Criticality Analysis Checklist

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

6. Curtis-Wright Flow Control and Services Corporation 10 CFR 2.390 Affidavit

7. NEDC-33932P, "Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis,"

Revision 1, dated October 2021 (Proprietary Version) cc: U.S. NRC Region III, Regional Administrator U.S. NRC Senior Resident Inspector, Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

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 2.0 DETAILED DESCRIPTION ................................................................................................ 2 2.1 Spent Fuel Pool Criticality Safety Analysis ..................................................................... 2 2.2 New Fuel Vault Criticality Safety Analysis ...................................................................... 3 2.3 Proposed Changes to Technical Specifications Section 4.3.1 ....................................... 3

3.0 TECHNICAL EVALUATION

............................................................................................... 4 3.1 Overview of System Design and Operation .................................................................... 4 3.2 Criticality Evaluation ....................................................................................................... 6 3.3 Accident Conditions ........................................................................................................ 7

4.0 REGULATORY EVALUATION

........................................................................................... 7 4.1 Applicable Regulatory Requirements/Criteria ................................................................. 7 4.2 Precedent ....................................................................................................................... 8 4.3 No Significant Hazards Consideration ............................................................................ 8 4.4 Conclusions .................................................................................................................. 11

5.0 ENVIRONMENTAL CONSIDERATION

........................................................................... 11

6.0 REFERENCES

................................................................................................................. 11 Page 1 of 12

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," Exelon Generation Company, LLC (EGC) requests an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, respectively. Specifically, EGC is utilizing a new criticality safety analysis (CSA) methodology (Reference 6.1) for performing the criticality safety evaluation for legacy fuel types in addition to the new GNF3 reload fuel design in the spent fuel pool (SFP). Use of the new SFP CSA methodology necessitates a change to the QCNPS Technical Specifications (TS) 4.3.1, "Criticality." EGC is also proposing a change to the new fuel vault (NFV) CSA to utilize the GESTAR II methodology (Reference 6.4, Section 3.5) 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 EGC intends to transition from the ATRIUM 10XM fuel design to the GNF3 fuel design at QCNPS beginning in the spring of 2023. The previous SFP CSAs (see References 6.7) were prepared by Holtec International Inc. (Holtec). The CSA for the QCNPS SFPs is now being rebaselined by GNF to:

Simplify the validation of GNF3 fuel designs against the CSA criteria. The new analysis will move QCNPS away from the need to validate the in-rack kinf value for each new lattice design to now validating the in-core 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.

Improve consistency among the Boiling Water Reactor (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 Technical Specifications.

The reason for this license amendment is the rebaselined 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 QCNPS Unit 1 and Unit 2 SFPs. The QCNPS Unit 1 and Unit 2 SFP racks are designed to accommodate BWR fuel. Both pools SFP racks credit Curtiss-Wrights NETCO-SNAP-IN rack inserts made of Boralcan. The SFP analysis does not credit any residual Boraflex material that may remain in the rack walls in the same manner as the previous NRC approved CSA for the introduction of rack inserts to the QCNPS SFPs (References 6.7). 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 regulatory limit of 0.95 at a 95 percent probability, 95 percent confidence level as required by 10 CFR 50.68(b) (e.g., QCNPS complies with the requirements specified in 10 CFR 50.68(b) instead of maintaining monitoring systems as described in 10 CFR 70.24). Reactivity effects of Page 2 of 12

ATTACHMENT 1 Evaluation of Proposed Changes abnormal and accident conditions are also evaluated to assure that under credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit.

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 6.2). Guidance pertaining to soluble boron in the SFP is not applicable because QCNPS 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). A copy of the NEI 12-16 Criticality Analysis Checklist is included as 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 necessitates a change to Technical Specifications 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 QCNPS NFV racks are GE designed low density racks with an interrack spacing of 11 inches (Reference 6.3, Section 9.1.1.2). The NFV rack CSA coverage for the new GNF3 fuel will be the GESTAR II (Reference 6.4) 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 6.6). The QCNPS NFV interrack pitch is 10.5 inches (the criteria listed in GESTAR II) and thus the racks may be utilized to store new GNF fuel with in-core SCCG kinf 1.31 (Reference 6.4, Section 3.5). Past NFV CSA will no longer be applicable to QCNPS 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 in-core SCCG limit of kinf 1.31 is the GESTAR II basis NFV CSA limit for QCNPS storage of fresh GNF3 fuel.

2.3 Proposed Changes to Technical Specifications Section 4.3.1 The QCNPS, Units 1 and 2 TS requirements related to spent fuel storage are contained in TS Section 4.3, "Fuel Storage." TS 4.3.1 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 6.22-inch center-to-center distance between fuel assemblies placed in the SFP storage racks in both pools. TS 4.3.1.1.d provides the limit on the areal density for the neutron absorber in the rack inserts as 0.0116 g10B/cm2. None of these items require update due to the proposed change in CSA methodology.

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ATTACHMENT 1 Evaluation of Proposed Changes 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.

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 1.29 gadolinia loading shall be limited to ensure fuel in the normal reactor core configuration at cold assemblies have a maximum k-infinity of conditions; and 0.8991 as determined at 4°C (39.2°F) in the normal spent fuel pool in-rack configuration; and 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 QCNPS Updated Final Safety Analysis Report (UFSAR) will be updated in accordance with 10 CFR 50.71(e) as part of implementation of the approved amendment. A summary of the proposed changes is provided below.

Section 9.1.1.2, Facilities Description, will be modified to reflect storage requirements of GNF3 fuel in the NFV, including receiving pre-channeled GNF3 fuel.

Section 9.1.1.3, Safety Evaluation, will be revised to reflect the NFV requirements of 10 CFR 50.68(b)(2) in this section and update cross-reference to ATRIUM 10XM with GNF3.

Sections 9.1.2.3, Safety Evaluation, and 9.1.2.3.1, "Safety Evaluation for Fuel," will be updated 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 in Section 9.1.

3.0 TECHNICAL EVALUATION

3.1 Overview of System Design and Operation The QCNPS UFSAR Section 9.1.2, "Spent Fuel Storage," documents the combined units SFP safety design bases as follows:

A. A maximum keff of 0.95 is maintained with the racks fully loaded with fuel of the highest anticipated reactivity and flooded with unborated water at a temperature corresponding to the highest reactivity. The criticality analyses include allowance for uncertainty and are described in a criticality analysis report applicable to the spent fuel pools.

B. The spent fuel storage racks, containing their full complement of fuel assemblies (i.e., 7554 fuel assemblies) are designed to withstand earthquake loadings of a Class 1 structure.

C. There will be no release of contamination or exposure of personnel to radiation in excess of 10 CFR 20 limits.

D. It is possible, at any time, to perform limited work on irradiated components.

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ATTACHMENT 1 Evaluation of Proposed Changes E. Pool storage space is provided for used control rods, flow channels, and other reactor components.

F. The fuel storage pools of Units 1 and 2 are connected by a double-gated transfer canal.

To achieve the safety design bases QCNPS has two joined SFPs which provide for storage of new unirradiated and irradiated fuel in a safe manner. The SFP facilities are designed to accept new unirradiated and irradiated fuel from both the Unit 1 and Unit 2 reactor cores (i.e., one units fuel may reside in either or both SFPs).

The QCNPS SFPs are identical in the types of SFP racks and neutron absorbing materials used. The SFPs contain high density spent fuel storage racks made up of 39 individual modules that have a combined capacity of 7554 fuel assemblies. The modules are a honeycomb arrangement of cells constructed of a series of cruciform shaped stainless steel elements. The high density spent fuel storage racks originally contained a nominal 0.070-inch-thick sheet of Boraflex neutron absorber material physically captured between the side walls of all adjacent boxes (cells). To provide space for the original neutron absorber sheet between each box wall, stainless steel spacer strips were placed between box wall plates.

The organic PDMS (polydimethylsiloxane) based Boraflex sheet material experienced premature degradation at QCNPS and across the industry. This was driven by high temperatures, high gamma radiation flux, and convection driven water flow that was able to enter and leave the areas between cells where the Boraflex resided. In response to the Boraflex degradation at QCNPS, all possible fuel storage cells in both SFPs racks had NETCO-SNAP-IN rack inserts installed (see Reference 6.7). A rack insert was installed in every rack cell location where a fuel bundle could be placed.

The rack inserts are made of a thin sheet of Rio Tinto Alcans Boralcan metal matrix composite material (formed from molten aluminum with a very fine particle B4C powder added) formed into a chevron shape that fully covers two of the interior sides of each rack cell in the axial range of the active fuel. All rack inserts were installed in the rack cells with the chevron corner in each cells south-west corner. In this way all fuel in the rack cells will have one Boralcan neutron absorber insert wing between them. The one exception is in the fuel rack cells along the SFPs north and east most racks edges. For these cells, the higher neutron radial leakage into the bulk pool water and surrounding structural materials helps offset the impact of having less neutron absorber. With the addition of the rack inserts, no negative reactivity credit is taken for residual Boraflex in the racks. The entire area that was originally occupied by Boraflex is now assumed in the CSA to contain water.

The specific NETCO-SNAP-IN rack inserts used at QCNPS have a minimum certified 10B areal density of 0.0116 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 Page 5 of 12

ATTACHMENT 1 Evaluation of Proposed Changes assembly accident conditions, and any uplift forces generated by the fuel handling equipment.

3.2 Criticality Evaluation In accordance with 10 CFR 50.68, a CSA for the QCNPS, Units 1 and 2 SFPs has been revised to support the purposes discussed in Section 2.1. The analysis, provided as Attachment 7, 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 NUREG-0800, Section 9.1.1 Revision 3 dated March 2007, have been met. NEI 12-16, Rev. 4 (Reference 6.2) was used as guidance for this analysis.

The revised CSA covers all legacy fuel in storage in either the QCNPS Unit 1 or Unit 2 SFP and the new GNF3 product line. A description of the GNF3 product line is provided in Section 4.1 of Attachment 7, while the disposition of legacy fuel is provided in Appendix B of Attachment 7.

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 Quad Cities 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 7 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 6.2). The CSA uses the minimum certified 10B areal density of 0.0116 g/cm2 in the Boralcan rack inserts at QCNPS.

The peak reactivity of the fuel in the QCNPS 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 7). The method of analysis is discussed in greater detail in Section 3.0 of Attachment 7. Validation of the codes and libraries is described in Section 3.4 and Appendix A of Attachment 7.

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ATTACHMENT 1 Evaluation of Proposed Changes The use of TGBLA06 for BWR core depletion calculations has been reviewed and accepted by the NRC as part of the approval of Reference 6.4. The NRC has also approved the MCNP-05P/TGBLA06 code package for use in similar fuel pool criticality analyses. Reference 6.5 documents an example of a previously NRC approved use of this code package.

3.3 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 Attachment 7. Note that the missing rack insert is conservatively treated as a normal condition bias in Section 5.5.2.

SFP temperature exceeding the normal range (moderator temperature/density changes)

Dropped and dropped + damaged fuel assemblies

Rack movement (seismic)

Mislocated fuel assembly (an assembly in the wrong location outside a storage rack)

The criticality analysis for the storage of BWR assemblies in the QCNPS SFP racks with Boralcan rack inserts 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 this bounding accident condition, i.e., the Dropped/Damaged Fuel (Case T14.B10), 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.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The regulations in 10 CFR 50.36, "Technical specifications," contain the requirements for the content of TSs. As required by 10 CFR 50.36(c)(4), "Design features," the Technical Specifications (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 paragraphs (c)(1), (2), and (3) of 10 CFR 50.36. QCNPS TS 4.0, "Design features", provides the QCNPS 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.

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ATTACHMENT 1 Evaluation of Proposed Changes 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 an 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 QCNPS NFV as it is a moderation controlled area (see Reference 6.3, Section 9.1.1.3). The QCNPS SFP criticality analysis, provided as Attachment 7 to this submittal along with the GESTAR II NFV criticality analysis in Reference 6.4, demonstrate that these requirements are met.

Paragraph (b)(7) of 10 CFR 50.68 states that the maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to 5.0 percent by weight. QCNPS GNF3 fuel is below 5.0 percent by weight 235U enrichment.

QCNPS, Units 1 and 2 were not licensed to the 10 CFR 50, Appendix A, General Design Criteria (GDC). The QCNPS, Units 1 and 2 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. 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 QCNPS's conformance with Criterion 66 is discussed in both Section 3.1.8.1, "Criterion 66 - Prevention of Fuel Storage Criticality" and Section 9.1, "Fuel and Storage Handling" of the QCNPS 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 QCNPS 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 has recently approved the use of the GNF CSA methodology to determine the acceptability of storing fresh and spent GNF3 fuel in other BWR spent fuel pools that contain NETCO-SNAP-IN rack inserts. These plants also utilized the NETCO-SNAP-IN rack inserts to provide fuel storage reactivity control in place of the degraded Boraflex material originally placed in the rack structure. One example of this is the acceptance of the CSA at Entergys River Bend Station as documented in the Reference 6.8 safety evaluation which was issued on December 31, 2019.

4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Page 8 of 12

ATTACHMENT 1 Evaluation of Proposed Changes Station (QCNPS), Units 1 and 2, respectively. Specifically, EGC is utilizing a new criticality safety analysis (CSA) methodology for evaluating the legacy fuel types and the new GNF3 reload fuel design in the spent fuel pool (SFP). Use of the new SFP CSA methodology necessitates a change to the QCNPS Technical Specifications (TS) 4.3.1, "Criticality." EGC 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 does not require a change to the QCNPS TS.

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

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

EGC has evaluated the proposed change for QCNPS using the criteria in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration.

The following information is provided to support a finding of no significant hazards consideration.

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed amendment involves a revised new fuel vault (NFV) criticality safety analysis (CSA) and a revised spent fuel pool (SFP) CSA for the QCNPS Units 1 and 2 spent fuel pools (SFPs) using new methodologies. Technical Specification 4.3.1.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 on the following four previously evaluated events and accidents was assessed:

A fuel handling accident (FHA),

A fuel mispositioning event,

A seismic event, and

A loss of SFP cooling event This 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.

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ATTACHMENT 1 Evaluation of Proposed Changes 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 fresh and spent fuel assemblies in the QCNPS, Units 1 and 2 shared NFV and SFPs is a normal activity for which QCNPS has been designed and licensed.

The proposed use of new methodologies for performing the QCNPS 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 pool, decay heat generation rate, or the SFP cooling and cleanup system.

The limiting dropped/damaged fuel event does not represent a new or different type of accident. Having a dropped/damaged fuel assembly within the fuel storage racks has always been possible, it was just not previously identified as a bounding type of event.

The associated analysis results show that the storage racks remain sub-critical following a worst-case dropped/damaged fuel event. Note that the missing rack insert 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 QCNPS 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. The 10 CFR 50.68(b)(2) regulation 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 Quad Cities 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 criticality analysis, that keff continues to be less than or equal to 0.95, thus preserving the required safety margin of 5 percent. The updated GNF methodology analysis, which also adds the GNF3 fuel type, results in less margin to the kmax(95/95) 0.95 regulatory limit (i.e., the new kmax(95/95) value is less than 0.009 closer to the limit than the current kmax(95/95) value). 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.

Page 10 of 12

ATTACHMENT 1 Evaluation of Proposed Changes Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, EGC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

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 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-33932, "Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis,"

Revision 0, dated May 2021 (Attachments 3 and 7 to RS-21-065 for the non-proprietary and proprietary versions, respectively) 6.2 NEI 12-16, Revision 4, "Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants," dated September 2019. (ADAMS Accession Number ML19269E069) 6.3 QCNPS Updated Final Safety Analysis Report (UFSAR), Revision 15, dated October 2019 6.4 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 ML20330A199 for the non-proprietary version)

Page 11 of 12

ATTACHMENT 1 Evaluation of Proposed Changes 6.5 Final Safety Evaluation for GE Hitachi Nuclear Energy Licensing Topical Report NEDC-33374P, Revision 3, "Safety Analysis Report for Fuel Storage Racks Criticality Analysis for ESBWR Plants," dated September 21, 2010 (ADAMS Accession Number ML102430580 for the non-proprietary version) 6.6 NEDC-33879P, Revision 4, GNF3 Generic Compliance with NEDE-24011-P-A (GESTAR II)," dated August 2020 (Enclosure to ADAMS Accession Number ML20244A104) 6.7 "Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments Regarding NETCO Inserts (TAC. NOS. MF2489 and MF2490) (RS-13-148)," dated December 31, 2014 (ADAMS Accession Number ML14346A306) 6.8 Final Safety Evaluation for River Bend Station, 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)

Page 12 of 12

ATTACHMENT 2 QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 MARK-UP OF QCNPS, UNITS 1 AND 2 TECHNICAL SPECIFICATIONS PAGES

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;

b. A nominal 6.22 inch center to center distance between fuel assemblies placed in the storage racks; Fuel assemblies having c. The combination of U-235 enrichment and gadolinia a maximum kinf of 1.29 loading shall be limited to ensure fuel assemblies in the normal reactor have a maximum k-infinity of 0.8991 as determined core configuration at at 4°C (39.2°F) in the normal spent fuel pool cold conditions in-rack configuration; and

d. The installed neutron absorbing rack inserts having a Boron-10 areal density 0.0116 g/cm2.

4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 666 ft 8.5 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 3657 fuel assemblies for Unit 1 and 3897 fuel assemblies for Unit 2.

Quad Cities 1 and 2 4.0-2 Amendment No. 253/248

ATTACHMENT 3 NEDC-33932, "Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis,"

Revision 1, dated October 2021 (Non-Proprietary Version)

Global Nuclear Fuel NEDO-33932 Revision 1 October 2021 Non-Proprietary Information Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis Copyright 2021 Global Nuclear Fuel - Americas, LLC All Rights Reserved

NEDO-33932 Revision 1 Non-Proprietary Information INFORMATION NOTICE This is a non-proprietary version of the document NEDC-33932P, Revision 1, 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 is furnished for the purpose of providing the results of the spent fuel pool criticality analysis for Quad Cities Nuclear Power Station, Units 1 and 2 (Quad Cities). The only undertakings of GNF with respect to information in this document are contained in the contracts between Exelon Generation Company (Exelon) and GNF, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Exelon, or for any purpose other than that for which it is intended 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-33932 Revision 1 Non-Proprietary Information Revision Status Revision Date Description of Change Number 0 May 2021 Initial release 1 October 2021 Revised marked proprietary content iii

NEDO-33932 Revision 1 Non-Proprietary Information Table of Contents 1.0 Introduction ........................................................................................................................1 2.0 Requirements......................................................................................................................2 3.0 Method of Analysis ............................................................................................................3 3.1 Cross-Sections................................................................................................................. 3 3.2 Geometry Treatment ....................................................................................................... 3 3.3 Convergence Checks ....................................................................................................... 4 3.4 Validation and Computational Basis .............................................................................. 4 3.5 In-Core k Methodology ................................................................................................. 7 3.6 Definitions....................................................................................................................... 9 3.7 Assumptions and Conservatisms .................................................................................... 9 4.0 Fuel Design Basis .............................................................................................................11 4.1 GNF3 Fuel Description ................................................................................................. 11 4.2 Fuel Model Description ................................................................................................ 15 5.0 Criticality Analysis of Spent Fuel Storage Racks .........................................................17 5.1 Description of Spent Fuel Storage Racks ..................................................................... 17 5.2 Spent Fuel Storage Rack Models .................................................................................. 18 5.3 Design Basis Lattice Selection...................................................................................... 20 5.4 Normal Configuration Analysis .................................................................................... 22 5.4.1 Analytical Models ......................................................................................................22 5.4.2 Results ........................................................................................................................23 5.5 Bias Cases ..................................................................................................................... 24 5.5.1 Depletion Bias Cases .................................................................................................24 5.5.2 Normal Bias Cases .....................................................................................................25 5.5.3 Abnormal/Accident Bias Cases .................................................................................26 5.5.4 Results ........................................................................................................................28 5.6 Uncertainty Analysis..................................................................................................... 30 5.6.1 Analytic Models .........................................................................................................30 5.6.2 Results ........................................................................................................................31 5.7 Maximum Reactivity .................................................................................................... 33 6.0 Conclusions .......................................................................................................................34 7.0 References .........................................................................................................................35 Appendix A - MCNP-05P Code Validation ...............................................................................36 A.1 - Trend Analysis ................................................................................................................. 41 A.2 - Bias and Bias Uncertainty Calculation - Single Sided Tolerance Limit ......................... 44 Appendix B - Legacy Fuel Storage Justification .......................................................................47 iv

NEDO-33932 Revision 1 Non-Proprietary Information List of Tables Table 1 - Summary kmax(95/95) Result ...........................................................................................1 Table 2 - Summary of the Critical Benchmark Experiments ..........................................................5 Table 3 - Area of Applicability Covered by Code Validation ........................................................6 Table 4 - Nominal Dimensions for GNF3 Fuel Lattice .................................................................13 Table 5 - Cell Dimensions.............................................................................................................13 Table 6 - Nominal Channel Dimensions for GNF3 Lattice ...........................................................14 Table 7 - GNF3 Fuel Stack Density as a Function of Gadolinia Concentration ...........................15 Table 8 - Storage Rack Dimensions ..............................................................................................19 Table 9 - Fuel Parameter Ranges Studied in Spent Fuel Rack......................................................21 Table 10 - Spent Fuel Storage Rack In-Rack k Results - Normal Configurations .....................23 Table 11 - Rack Periphery Study Results......................................................................................25 Table 12 - Results for a Misplaced Bundle ...................................................................................27 Table 13 - Spent Fuel Storage Rack Abnormal Bias Summary ....................................................28 Table 14 - Spent Fuel Storage Rack Bias Summary .....................................................................29 Table 15 - Spent Fuel Storage Rack Tolerance and Uncertainty k Values ................................32 Table 16 - Spent Fuel Storage Rack Results Summary ................................................................33 Table 17 - MCNP-05P Results for the Benchmark Calculations ..................................................36 Table 18 - Trending Parameters ....................................................................................................41 Table 19 - Trending Results Summary .........................................................................................44 Table 20 - Bias and Bias Uncertainty for MCNP-05P with ENDF/B-VII ....................................46 Table 21 - Recommended Bias and Bias Uncertainty in Criticality Analyses for MCNP-05P with ENDF/B-VII ....................................................................................................................46 Table 22 - Limiting Cold As-Designed Eigenvalue of Bundles Inserted Into Quad Cities ..........47 Table 23 - Legacy Bundles with Missing Rod Locations at Quad Cities .....................................48 v

NEDO-33932 Revision 1 Non-Proprietary Information List of Figures Figure 1 - GNF3 Lattice Configuration .........................................................................................12 Figure 2 - GNF3 Channel Dimensions..........................................................................................14 Figure 3 - GNF3 MID Lattice in MCNP-05P ...............................................................................16 Figure 4 -Spent Rack Array Without Inserts ..................................................................................17 Figure 5 - Storage Rack Model Schematic....................................................................................18 Figure 6 - Zoomed Storage Rack Model Schematic .....................................................................19 Figure 7 - Spent Fuel In-Core versus In-Rack Eigenvalues ..........................................................22 Figure 8 - Finite Misplaced Bundle Model Example ....................................................................27 Figure 9 - Scatterplot of knorm versus EALF ..................................................................................41 Figure 10 - Scatterplot of knorm versus 235U wt.%{3} .....................................................................42 Figure 11 - Scatterplot of knorm versus 239Pu wt.% ........................................................................42 Figure 12 - Scatterplot of knorm versus H/X ...................................................................................43 Figure 13 - Normality Test of knorm Results ..................................................................................45 vi

NEDO-33932 Revision 1 Non-Proprietary Information ACRONYMS Term Definition 2D Two-Dimensional AOA Area of Applicability BAF Bottom of Active Fuel BASE Base Lattice BOL Beginning-of-Life BWR Boiling Water Reactor CFR Code of Federal Regulations CW Curtiss-Wright Flow Control Service, LLC EALF Energy of the Average Lethargy Causing Fission

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

GEH GE-Hitachi Nuclear Energy Americas LLC GNF Global Nuclear Fuel - Americas, LLC HTC Haut Taux de Combustion H/X Hydrogen-to-Fissile Ratio MID Mid Lattice MOX Mixed Uranium-Plutonium Oxide NCA Nuclear Critical Assembly NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission SCCG Standard Cold Core Geometry SS Stainless Steel VAN Vanished Lattice WREC Westinghouse Reactor Evaluation Center UO2 Uranium Dioxide vii

NEDO-33932 Revision 1 Non-Proprietary Information

1.0 INTRODUCTION

This report describes the criticality analysis and results for the Quad Cities Boraflex spent fuel racks with credit for NETCO-SNAP-IN neutron absorbing inserts. No credit for the Boraflex neutron absorber is taken in this analysis. The methodology and analytical models utilized in this criticality analysis confirm that the storage rack systems have been accurately and conservatively represented. This analysis covers the future GNF3 fuel product designs and all legacy fuel stored in Quad Cities spent fuel pools.

The racks are analyzed using the MCNP-05P Monte Carlo neutron transport program and ENDF/B-VII.0 cross-section library. The methodology used in this analysis is the peak Standard Cold Core Geometry (SCCG) in-core eigenvalue (k) criterion methodology. A maximum SCCG, uncontrolled peak in-core k of 1.29 as defined by the lattice physics code TGBLA06 (Reference

1) is set as the limit for this analysis. As demonstrated in Table 1, the analysis resulted in a storage rack maximum k-effective (kmax(95/95)) less than 0.95 for normal and credible abnormal operation with tolerances and uncertainties taken into account.

Table 1 - Summary kmax(95/95) Result Region kmax(95/95)

Spent Fuel Pool Racks 0.94200 1

NEDO-33932 Revision 1 Non-Proprietary Information 2.0 REQUIREMENTS Title 10 of the Code of Federal Regulations (CFR) Part 50 defines the requirements for the prevention of criticality in fuel storage and handling at nuclear power plants. 10 CFR 50.68 details specifically that the storage rack kmax(95/95) for spent fuel storage racks must be demonstrated to be 0.95 for normal and credible abnormal operation with tolerances and computational uncertainties taken into account. The Standard Review Plan (Reference 2) outlines the standards that must be met for these analyses. All necessary requirements are met in this analysis. Nuclear Energy Institute (NEI) 12-16 (Reference 3) is used as the guidance document for this analysis.

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NEDO-33932 Revision 1 Non-Proprietary Information 3.0 METHOD OF ANALYSIS In this evaluation, in-core k values and exposure dependent, pin-by-pin isotopic specifications are generated using the GE-Hitachi Nuclear Energy Americas LLC (GEH)/GNF lattice physics production code TGBLA06. TGBLA06 solves Two-Dimensional (2D) diffusion equations with diffusion parameters corrected by transport theory to provide system multiplication factors and perform burnup calculations.

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 is 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 point-wise (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 an 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.

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NEDO-33932 Revision 1 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 and more total cycles in order to meet all the converge 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/CR-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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NEDO-33932 Revision 1 Non-Proprietary Information Table 2 - Summary of the Critical Benchmark Experiments Experiment Experiments Year Where

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NEDO-33932 Revision 1 Non-Proprietary Information Table 3 - Area of Applicability Covered by Code Validation Validation Spent Fuel Rack Parameters Area of Applicability Characteristics Fissionable Material Uranium, Plutonium Uranium, Actinides Chemical Form UO2, MOX UO2, MOX Enrichment (wt.% 235U) wt.% 235U 4.9 wt.% 235U 4.9 Enrichment (wt.% 239Pu) wt.% 239Pu 5.3 wt.% 239Pu 4.9 Physical Form Solid Compound Solid Compound Temperature ~20°C up to ~100°C 4-126°C Moderator (in fuel region) H2O H2O Physical Form Solution Solution Temperature ~20°C up to ~100°C 4-126°C Reflector (in fuel region) H2O H2O Physical Form Solution Solution Temperature 20°C 4-126°C None/Boron/Gadolinium Boron/Gadolinium/

Absorbers Stainless Steel (SS)/Copper Fission Products Neutron Energy Spectrum Thermal Thermal Energy of Average Lethargy 3.666E-07 6.8E 8.6E-7 Causing Fission (MeV) (Limiting In-rack k Case)

Table 3 demonstrates that the AOA of this validation encompasses the majority of storage characteristics of new fuel in the spent fuel storage racks. ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

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For the storage of spent fuel, however, it is appropriate to add additional uncertainty terms to the kmax(95/95) result. Specifically, these items are:

Uncertainty in fuel depletion calculations Consistent with NEI 12-16, a conservative approximation of the fuel depletion uncertainty was quantified by assessing the reactivity difference between a Beginning-of-Life (BOL) system and the exposure dependent, peak reactivity system of interest. Specifically, the cold, in-core, BOL reactivity of the spent fuel rack design basis bundle with no gadolinium present was compared to the reactivity of the exposed design basis bundle at its cold, in-core, peak reactivity statepoint. Both reactivities are calculated for comparison in the rack 6

NEDO-33932 Revision 1 Non-Proprietary Information system. Five percent of the difference in reactivities between these two cases is included as an uncertainty to the spent fuel rack studies in Table 15 to cover the depletion isotopic benchmarking gap including gap for minor actinides and fission products.

TGBLA06 eigenvalue uncertainty An additional uncertainty is also added to the fuel rack studies 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 peak 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 a number of plants. This report demonstrates that the methodology is also appropriate for use at Quad Cities 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 Quad Cities fuel storage racks is provided below.

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NEDO-33932 Revision 1 Non-Proprietary Information

1. Identify the unique lattices in each assembly design.

2. Deplete the lattices in TGBLA06 using the following conditions:

a. Centered Assembly according to Quad Cities specific lattice spacing and zero leakage

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3. Ensure that the k values obtained from Step 3 for each lattice are less than or equal to the k limit of 1.29.

Documentation that all legacy fuel types at Quad Cities currently comply with this in-core limit is documented in Appendix B.

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NEDO-33932 Revision 1 Non-Proprietary Information 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 (BASE) - An axial zone of a fuel assembly typically 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.

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. ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

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2. ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

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NEDO-33932 Revision 1 Non-Proprietary Information

`````` ```````````````````````````````````````````````````

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3. Design basis lattices with in-core k values greater than the proposed 1.29 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. ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

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` ` ` ))

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).

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 inserts. Each insert is assumed to contain the minimum areal density of 0.0116 g 10B/cm2. All other insert material is ignored. Ignoring the other materials conservatively limits neutron absorption in the insert.

10. No credit is taken for the Boraflex in the storage racks in the analysis, and all material between the inner cell walls is modeled as water. Modeling this material as water is reasonable, as there is not a water tight seal between the Boraflex and pool environment, and therefore any significant gap formations within the poison material will be filled with water.

11. Each Boraflex rack cell will contain one chevron-shaped insert, and the inserts will be oriented uniformly in all the Boraflex racks. This analysis assumes the two neutron absorber panels in each chevron-shaped insert in the rack cells will be oriented such that one panel will be on the south side and the other panel will be on the west side. In this orientation there will be no insert absorber material between the outer edges of the Boraflex racks and the north and east pool walls.

10

NEDO-33932 Revision 1 Non-Proprietary Information 4.0 FUEL DESIGN BASIS This rack criticality analysis covers all legacy and current fuel in Quad Cities, and the planned GNF3 future fuel product line. The disposition for all legacy and current fuel is in Appendix B.

The description of the GNF3 fuel product lines is found in Section 4.1. This product line 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 are explicitly assessed in Appendix B.

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 12

NEDO-33932 Revision 1 Non-Proprietary Information Table 4 - Nominal Dimensions for GNF3 Fuel Lattice Dimension Item mm in

((` ` ` ` ` ` ` ` ` ` ` ` ` ````` `````

Channel

```````````````````` ` ````` `````

`````````````` ` ``` ``````

Fuel Rod ``````````````````` ` ````` ``````

``````````````````` ` ```` ``````

Pellet `````````````` ` ```` ` ` ` ` ` ` ` ` ` ` ))

`````````````

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `````````````` ` `````````````

``````````````

``` `````````````` ` `````````````

` ` ` ))

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ````` `````

Bundle Lattice `````````````` ` ```` `````

``````````````````````` ` ````` ` ` ` ` ` ` ` ` ` ))

Table 5 - Cell Dimensions Lattice Channel 1/2 Wide Gap, Q 1/2 Narrow Gap, R Control Blade Pitch, S Type Name mm in mm in mm in

((` ```` ```` ````` ```` ````` `````` ` ` ` ` ` ` ` ` ` ` ))

13

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((

` ` ` ` ` ))

Figure 2 - GNF3 Channel Dimensions Table 6 - Nominal Channel Dimensions for GNF3 Lattice Channel Name 83AV 93AV Channel Section Zone 1 Zone 2 Zone 1 Zone 2 Dimension mm in mm in mm in mm in

((` ` ` ` ` ` ` ` ` ` ` `

``` ```` ````` ```` ````` ```` ````` ```` `````

````

`````````````

``` `` `` ```` ````` `` `` ```` `````

`

```````````` ``` `` `` ````` ````` `` `` ````` `````

````````````` `````````

``` `` `` ```` ````` `` `` ````

```` ))

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NEDO-33932 Revision 1 Non-Proprietary Information Table 7 - GNF3 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 conditions ((` ` ` ` ` ` ` ` ` ` ` `

```````````````````````````````````````````````` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) 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 concentration, and void histories considered to be reasonably representative of any Quad Cities fuel designs. 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.

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NEDO-33932 Revision 1 Non-Proprietary Information 5.0 CRITICALITY ANALYSIS OF SPENT FUEL STORAGE RACKS 5.1 Description of Spent Fuel Storage Racks The Quad Cities Boraflex storage racks manufactured by Joseph Oat Corporation consist of multiple modular cruciform, T-shaped, and L-shaped 304 SS structures that form rack cells with a center-to-center cell pitch of 6.22 inches. These structures contain 0.070-inch thick Boraflex panels sandwiched between the 0.075-inch SS outer walls of the modular shapes. A schematic of a Boraflex storage rack array without inserts installed is shown in Figure 4.

Figure 4 -Spent Rack Array Without Inserts Originally, the racks employed thermal neutron absorption in the 10B of the Boraflex as the primary mechanism of reactivity control; however, the Boraflex has been demonstrated to be degrading over time. Therefore, no credit is taken for the Boraflex in this analysis, and all material between the inner cell wall and outer wrapper is modeled as water. Modeling this material as water is reasonable, as the outer wrapper does not provide a water tight seal between the Boraflex and pool environment.

Therefore, any significant gap formations within the poison material will be filled with water.

To supplement the reactivity suppression capability of the rack, chevron shaped neutron absorbing inserts (NETCO-SNAP-IN) are installed in each of the storage cells in a storage rack module. These inserts extend over the full-length of the active fuel region of the storage assemblies. The inserts are manufactured from a Rio Tinto Alcan aluminum boron carbide metal matrix composite with a minimum certified areal density of 0.0116 g 10B/cm2. The nominal designed wing length of the inserts is (( )) inches, and the nominal thickness is 0.085 inches. Each insert is installed with the same orientation. In this way, one leg of an insert exists between each bundle in the storage rack assembly.

Figures 5 and 6 demonstrate where the inserts are located in each cell.

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NEDO-33932 Revision 1 Non-Proprietary Information Based on the insert configuration, peripheral storage cells on the north and east sides of the storage pools will not be surrounded by four wings of the absorbing insert. The reactivity effect of this storage limitation is assessed in Section 5.5.

5.2 Spent Fuel Storage Rack Models This analysis covers a single bounding storage configuration of maximum reactivity fuel in every storage location with a NETCO-SNAP-IN insert in every storage location.

A 2D infinite storage array with periodic boundary conditions is modeled to conservatively represent the nominal spent fuel pool configuration. An image of a single element of the model is provided in Figure 5 and a zoomed in view of Figure 6, with dimensions and tolerances presented in Table 8. This single element is used to define a 10x10 rack array with periodic boundary conditions. This array is used in the design basis bundle selection process in Section 5.3.

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

((

` ` ` ` ` ))

Figure 5 - Storage Rack Model Schematic 18

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((

` ` ` ` ` ))

Figure 6 - Zoomed Storage Rack Model Schematic Table 8 - Storage Rack Dimensions Tolerances Rack Model Parameter Nominal Plus Minus (inches) (inches) (inches)

Rack Cell Inner Dimension 6.00 0.125 0.000 Rack Pitch 6.22 0.125 0.000 Inner Cell Wall Thickness 0.075 0.004 0.004 Boraflex Thickness 0.070 0.007 0.007 Boraflex Width 5.80 - -

Rack Insert Width (( ))

Rack Insert Thickness 0.085 0.005 0.005 19

NEDO-33932 Revision 1 Non-Proprietary Information 5.3 Design Basis Lattice Selection Table 9 defines the lattice designs and exposure histories that were explicitly studied in the spent fuel storage rack to determine the geometric configuration and isotopic composition that results in the worst rack efficiency. Note that void state is not a relevant parameter for zero exposure peak reactivity cases, and, therefore, only a single result is presented for these fuel loadings. Figure 7 presents a graph that demonstrates the linear nature of the in-core to in-rack results over all rack efficiency cases studied in the rack system. This figure also provides infinite in-core and in-rack eigenvalue pairs ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) to allow for the linear relationship to be demonstrated over a large range of exposures. The highest rack efficiency with an in-core k greater than the proposed limit of 1.29 is found to result from the parameters defined in Table 9 Case 12. The geometry and isotopics defined for this case are used to define all bundles in the remaining spent fuel rack analyses.

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NEDO-33932 Revision 1 Non-Proprietary Information Table 9 - Fuel Parameter Ranges Studied in Spent Fuel Rack Average Number Peak-Lattice Lattice Gadolinia Reactivity TGBLA06 MCNP-05P of Defined Defined Rack Case Type Void Enrichment Gadolinia Concentration Exposure Efficiency (Gd wt. %) In-Core k In-Rack k (235U wt.%) Rods (GWD/ST) 1 ((` ` ` ` ` ` ` ` ` ` ```` ` ` ` ``````` 0.87926 ((` ` ` ` ` ` `

2 ````````` ` ``` ` ` ` ``````` 0.86289 ```````

3 ````````` ` ``` `` ` `` ``````` 0.90397 ```````

4 ````````` `` ``` `` ` `` ``````` 0.89529 ```````

5 ````````` `` ``` `` ` `` ``````` 0.88354 ```````

6 ```````` ` ```` ` ` ` ``````` 0.90023 ```````

7 ```````` `` ```` ` ` ` ``````` 0.89747 ```````

8 ```````` `` ```` ` ` ` ``````` 0.89184 ```````

9 ```````` ` ``` ` ` ` ``````` 0.88654 ```````

10 ```````` `` ``` ` ` ` ``````` 0.88511 ```````

11 ```````` `` ``` ` ` ` ``````` 0.88231 ```````

12 ```````` ` ``` `` ` `` ``````` 0.91396 ```````

13 ```````` `` ``` `` ` `` ``````` 0.90812 ```````

14 ```````` `` ``` `` ` `` ``````` 0.89816 ```````

15 ```````` ` ``` `` ` `` ``````` 0.90639 ```````

16 ```````` `` ``` `` ` `` ``````` 0.90248 ```````

17 ```````` `` ``` `` ` `` ``````` 0.89492 ```````

18 ```````` ` ```` ` ` ` ``````` 0.87694 ```````

``````````` ```````````

19 ```````` ` ```` ` ` ` 0.86723

)) ))

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NEDO-33932 Revision 1 Non-Proprietary Information

((

` ` ` ` ` ))

Figure 7 - Spent Fuel In-Core versus In-Rack 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 for this analysis consistent with NEI 12-16 (Reference 3).

((` ` ` ` ` ` ` ` ` ` ` ` ` ` `

``````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

Pool moderator temperature variation As the non-channeled assembly evaluation demonstrated a decrease in reactivity when compared to nominal, channeled storage conditions, the studies are performed with channeled bundles.

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NEDO-33932 Revision 1 Non-Proprietary Information 5.4.2 Results The results of the study are provided in Table 10. This information demonstrates that none of the normal configurations analyzed increase the system reactivity by a statistically significant amount over the nominal loading pattern. The in-rack k associated with this nominal combination of conditions is 0.91396, and is hereafter referred to as kNormal. This configuration will be used for all abnormal and tolerance studies that are performed on an infinite basis. Any small, positive reactivity differences from this nominal condition are included in the calculation of the system bias in Section 5.5.4.

Table 10 - Spent Fuel Storage Rack In-Rack k Results - Normal Configurations MCNP-05P In-Rack Term Configuration Uncertainty k

(1)

Base Nominal - Centered, channeled, ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) 0.91396 ((` ` ` ` ` ` `

kN1 Non-channeled assemblies 0.91075 ```````

kN2 ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` 0.91471* ```````

kN2 ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) 0.91448 ```````

kN3 Moderator Temperature decrease to 4oC (=1 g/cc) 0.91511* ```````

Moderator Temperature increase to 126oC with 20% void kN3

(=0.7508 g/cc) 0.87609 ` ` ` ` ` ` ` ` ` ` ` ))

Largest positive reactivity increase from nominal case for each term is included in roll-up of kBias 23

NEDO-33932 Revision 1 Non-Proprietary Information 5.5 Bias Cases 5.5.1 Depletion Bias Cases The following configurations related to the depletion conditions of the stored bundles were explicitly considered, where each description defines a condition all bundles in storage experience over their entire exposure histories. These bound the conditions the bundles actually experience.

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

```````````````````````````````````````````````

``````````````````````````````````````````````````````

`````````````````````````````````````````````

``````````````````````````````````````````````

`````````````````````````````````````````````````````````````````````````````````````````

````````````````````````````````

`````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

Depleted with clad creep 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

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

`````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

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NEDO-33932 Revision 1 Non-Proprietary Information 5.5.2 Normal Bias Cases The following bias cases are included for normal conditions. As seen in Table 10, cases with a moderator temperature decrease ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) resulted in the largest positive reactivity increases from the nominal case for their respective terms and are therefore included in Table 14.

No inserts on rack periphery As discussed in Section 5.1 and illustrated in Figure 5, there will be assemblies loaded in storage cells on two sides that will not be surrounded by neutron absorbing inserts.

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

`````````````````````````````````````````````````````````````````````````````````````````

` ` ` )) Results are provided in Table 11. The reactivity increase from this study is included in the final kBias term in Table 14.

Table 11 - Rack Periphery Study Results MCNP-05P Description keff Uncertainty k (1)

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ``````` ``````` ` ` ` ` ))

No Inserts on Rack Periphery ((` ` ` ` ` ` ` ``````` ` ` ` ` ` ` ` ` ` ` ` ))

Missing rack insert A missing insert from the 10x10 infinite array was analyzed to cover the periodic removal of an insert for inspection or an insert being accidently removed during fuel movement.

The relative reactivity increase from this condition is included in the bias table in Table 14.

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NEDO-33932 Revision 1 Non-Proprietary Information 5.5.3 Abnormal/Accident Bias Cases Additionally, perturbations of the normal spent fuel rack configuration were considered for credible accident scenarios. The scenarios considered are presented in the bulleted lists that follow, with explanations of the abnormal condition provided below each listing of similar configurations.

The most limiting of these abnormal conditions is included in the final kBias term in Table 14.

Dropped/damaged fuel Justification - The dropped/damaged fuel scenario ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

`````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) The relative reactivity change from this abnormal condition is included in Table 14.

Abnormal positioning of a fuel assembly outside the fuel storage rack Justification - There is enough space for an abnormally positioned bundle between:

o the south, east, and west sides of the Boraflex racks and the pool wall, o the north side one of the Boraflex racks and the dry cask storage pad, and o between the north side of one of the Boraflex racks and an adjacent Boraflex rack in the spent fuel pool.

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

`````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

A misplaced bundle outside the rack is analyzed on an edge of the rack ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

`````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) The calculation was then reperformed several times with a misplaced bundle oriented flush with a rack ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) the most limiting result generated is used to determine the k from the base case eigenvalue as shown in Table 12. The most limiting orientation identified in the study ((` ` ` ` ` ` ` ` ` ` ` ` `

`````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) depicted in Figure 8.

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((

` ` ` ` ` ))

Figure 8 - Finite Misplaced Bundle Model Example Table 12 - Results for a Misplaced Bundle MCNP-05P Description keff Uncertainty k (1)

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ``````` ``````` ` ` ` ` ` ))

Misplaced Bundle, in the most limiting location (Figure 8) ((` ` ` ` ` ` ` ``````` ` ` ` ` ` ` ` ` ` ` ` ))

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NEDO-33932 Revision 1 Non-Proprietary Information The following abnormal configurations are also considered bounded, with the justification provided:

Dropped bundle on rack Justification - For a drop on the rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the fuel in the rack of more than 12 inches. At this separation distance, the fissile material will be separated by enough neutron mean free paths to preclude neutron interactions that increase keff, and the overall effect on reactivity will be insignificant.

Rack sliding due to seismic event which causes water gap between racks to close Justification - 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 Justification - 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.

Table 13 - Spent Fuel Storage Rack Abnormal Bias Summary MCNP-05P k Uncertainty Description keff Uncertainty k (2)

(1)

Dropped/Damaged Fuel 0.91498 ((` ` ` ` ` ` ` ` ` ` ` )) 0.00102 ((` ` ` ` ` ` ` ` ` ` ` ))

Misplaced Bundle* ((` ` ` ` ` ` ` ``````` ``````` ` ` ` ` ` ` ` ` ` ` ` ))

Per the double contingency principle (Reference 3), only the most limiting misplaced bundle case is included in the bias roll-up in Table 14.

5.5.4 Results The results of the abnormal studies are provided in Table 14. The k term in this table represents the difference between the system reactivity with the specified bias case and kNormal for terms kB1 through kB5. (( ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) kB6 and kB7 are the normal condition cases that resulted in positive reactivity contributions. kB8 is extracted from Table 11. kB9 is the missing insert case and kB10 and kB11 are extracted from Table 13. 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, a kBi value must be both positive and the largest for its respective term to be considered.

n k Bias = k Bi (1) i =1 28

NEDO-33932 Revision 1 Non-Proprietary Information Table 14 - Spent Fuel Storage Rack Bias Summary MCNP-05P k Uncertainty Term Description keff Uncertainty k*

(2)

(1)

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

kB1 0.91264 ((` ` ` ` ` ` ` -0.00132 ((` ` ` ` ` ` `

`

kB2 ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` 0.91549 ``````` 0.00153 ```````

```````````````````````````````````

kB2 0.91517 ``````` 0.00121 ```````

`````

kB3 ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` 0.91485 ``````` 0.00089 ```````

kB3 ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` 0.91480 ``````` 0.00084 ```````

```````````````````````````````````

kB4 0.91494 ``````` 0.00098 ```````

````

```````````````````````````````````

kB4 0.91466 ``````` 0.00070 ```````

` ` ` ` ` ` ` ` ` ))

kB5 Depleted with clad creep 0.91526 ``````` 0.00130 ```````

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

kB6 0.91471 ``````` 0.00075 ```````

` ))

Moderator Temperature decrease to kB7 0.91511 ` ` ` ` ` ` ` ` ` ` ` )) 0.00115 ` ` ` ` ` ` ` ` ` ` ` ))

4oC (=1 g/cc) kB8 No inserts on rack periphery ((` ` ` ` ` ` ` ``````` ``````` ` ` ` ` ` ` ` ` ` ` ` ))

kB9 Missing insert 0.91853 ((` ` ` ` ` ` ` 0.00457 ((` ` ` ` ` ` `

kB10 Dropped/Damaged Fuel 0.91498 ` ` ` ` ` ` ` ` ` ` ` )) 0.00102 ` ` ` ` ` ` ` ` ` ` ` ))

kB11 Misplaced Bundle ((` ` ` ` ` ` ` ``````` ``````` ` ` ` ` ` ` ` ` ` ` ` ))

kBias ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

For conservatism, only positive values that are the largest for their respective term are considered.

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

29

NEDO-33932 Revision 1 Non-Proprietary Information 5.6 Uncertainty Analysis 5.6.1 Analytic Models The following tolerance study configurations were explicitly considered for the spent fuel rack:

Fuel enrichment increases by ((` ` ` ` ` ` ` ` ` ` ` ` ` ` )) 235U

Fuel pellet density increased by ((` ` ` ` ` ` ` ` ` )) of nominal value

Gadolinia concentration decreased by ((` ` ` ` ` ` ` ` ` ` ` ` ))

Rod cladding thickness increased by ((` ` ` ` ` ` ` )) and rod cladding outer diameter increased by ((` ` ` ` ` ` ))

Rod cladding thickness decreased by ((` ` ` ` ` ` ` )) and rod cladding outer diameter decreased 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 decrease by 0.004 inches

Rack wall thickness increase by 0.004 inches

Rack pitch increase by 0.125 inches

Rack insert thickness decrease by 0.005 inches

Rack insert thickness increase by 0.005 inches

Rack insert width decrease by (( ))

Rack insert width increase by (( ))

All the tolerances used in these analyses are at least 2 design limits. The models developed for these studies were all based on the normal configuration presented in Section 5.4.

There is no manufacturing tolerance for a decrease in rack pitch; therefore, no tolerance case was performed on the pitch decrease.

Because the Boraflex is modeled as water in this analysis, no tolerance cases are performed on the Boraflex thickness or width.

10 This analysis uses the certified minimum B areal density; therefore, no tolerance case was performed on the insert 10B density.

30

NEDO-33932 Revision 1 Non-Proprietary Information 5.6.2 Results The results of the tolerance studies and uncertainties are provided in Table 15. The values are summed using Equation 2, which is adopted from NEI 12-16 (Reference 3). The kTi terms in this table represent 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.

n n kUncertainty = k + k i =1 2

Ti i =1 2

Ui (2) 31

NEDO-33932 Revision 1 Non-Proprietary Information Table 15 - Spent Fuel Storage Rack Tolerance and Uncertainty k Values MCNP-05P k Uncertainty Term Description keff k Uncertainty (1) (2)+

kT1 Fuel enrichment increase 0.91814 ((` ` ` ` ` ` ` 0.00418 ((` ` ` ` ` ` `

kT2 Fuel pellet density increase 0.91529 ``````` 0.00133 ```````

kT3 Gadolinia wt.% decrease 0.91998 ``````` 0.00602 ```````

Rod clad thickness/outer diameter kT4 0.90908 ``````` -0.00488 ```````

increase Rod clad thickness/outer diameter kT4 0.91990 ``````` 0.00594 ```````

decrease kT5 Channel thickness increase 0.91489 ``````` 0.00093* ```````

kT5 Channel thickness decrease 0.91475 ``````` 0.00079 ```````

kT6 Pellet outer diameter increase 0.91504 ``````` 0.00108* ```````

kT6 Pellet outer diameter decrease 0.91378 ``````` -0.00018 ```````

kT7 Fuel rod pin pitch increase 0.91618 ``````` 0.00222* ```````

kT7 Fuel rod pin pitch decrease 0.91334 ``````` -0.00062 ```````

kT8 Rack wall thickness increase 0.91476 ``````` 0.00080* ```````

kT8 Rack wall thickness decrease 0.91463 ``````` 0.00067 ```````

kT9 Rack pitch increase 0.89850 ``````` -0.01546 ```````

kT10 Rack insert thickness decrease 0.91446 ``````` 0.00050 ```````

kT10 Rack insert thickness increase 0.91554 ``````` 0.00158* ```````

kT11 Rack insert width decrease 0.91553 ``````` 0.00157* ```````

kT11 Rack insert width increase 0.91452 ` ` ` ` ` ` ` ` ` ` ` )) 0.00056 ` ` ` ` ` ` ` ` ` ` ` ))

Critical benchmark bias uncertainty kU1 (95/95) (MCNP-05P versus critical ((` ` ``````` `

experiments) kU2 TGBLA06 eigenvalue uncertainty (95/95) ` ` ``````` ` ` ` ` ` ))

Uncertainty on kNormal (2 x 1 value for kU3 - ((` ``````` `

base term in Table 10) kU4 Uncertainty of k bias contributors (2) - ` ``````` `

Uncertainty of k tolerance contributors kU5 - ` ``````` `

(2) kU6 Uncertainty in fuel depletion - ` ``````` ` ` ` ` ` ))

kUncertainty ((` ` ` ` ` ` ` ` ` ` ` ))

For conservatism, only positive values that are the largest for their respective term are considered.

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

32

NEDO-33932 Revision 1 Non-Proprietary Information 5.7 Maximum Reactivity The maximum reactivity of the spent fuel rack without crediting Boraflex and with rack inserts installed, considering all biases, tolerances, and uncertainties, is calculated using Equation 3. The final values are presented in Table 16.

k max ( 95 / 95 ) = k Normal + k Bias + k Uncertaint y (3)

Table 16 - Spent Fuel Storage Rack Results Summary Term Value kNormal 0.91396 kBias ((` ` ` ` ` ` `

kUncertainty ` ` ` ` ` ` ` ` ` ` ` ))

kmax(95/95) 0.94200

((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ))

33

NEDO-33932 Revision 1 Non-Proprietary Information

6.0 CONCLUSION

S The Quad Cities 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 1.29 as defined by TGBLA06 is specified as the rack design limit for GNF3 fuel in the spent fuel racks with NETCO-SNAP-IN rack inserts installed. 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. Documentation that all legacy Quad Cities fuel types meet the kmax(95/95) limit is found in Appendix B.

34

NEDO-33932 Revision 1 Non-Proprietary Information

7.0 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 Revision 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants, September 2019. (NRC ADAMS Accession Number ML18088B400).

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.

35

NEDO-33932 Revision 1 Non-Proprietary Information APPENDIX A - MCNP-05P CODE VALIDATION Table 17 presents the results of the benchmark calculations described in Section 3.4. Note that it is necessary to make an adjustment to the calculated keff value if the critical experiment being modeled was not at a critical state. This adjustment is done by normalizing the kcalc values to the experimental values, which is valid for small differences in keff. This normalization is reported as knorm and is determined using Equation A-1. The combined uncertainty ( ) from the measurement and the calculation is also determined using Equation A-2.

= / (A-1)

= + (A-2)

Table 17 - MCNP-05P Results for the Benchmark Calculations Benchmark Experimental MCNP-05P MCNP-05P Norm. Combined Expt. Eigenvalue Uncertainty Result Result

Experiment Uncertainty Uncertainty (kexp) (exp) (kcalc) (calc) (knorm) (t)

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36

NEDO-33932 Revision 1 Non-Proprietary Information Benchmark Experimental MCNP-05P MCNP-05P Norm. Combined Expt. Eigenvalue Uncertainty Result Result

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37

NEDO-33932 Revision 1 Non-Proprietary Information Benchmark Experimental MCNP-05P MCNP-05P Norm. Combined Expt. Eigenvalue Uncertainty Result Result

Experiment Uncertainty Uncertainty (kexp) (exp) (kcalc) (calc) (knorm) (t)

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38

NEDO-33932 Revision 1 Non-Proprietary Information Benchmark Experimental MCNP-05P MCNP-05P Norm. Combined Expt. Eigenvalue Uncertainty Result Result

Experiment Uncertainty Uncertainty (kexp) (exp) (kcalc) (calc) (knorm) (t)

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39

NEDO-33932 Revision 1 Non-Proprietary Information Benchmark Experimental MCNP-05P MCNP-05P Norm. Combined Expt. Eigenvalue Uncertainty Result Result

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40

NEDO-33932 Revision 1 Non-Proprietary Information A.1 - Trend Analysis To determine if any trend is evident in this pool of experiments, the parameters listed in Table 18 were considered as independent variables.

Table 18 - Trending Parameters Energy of the Average Lethargy causing Fission (EALF)

Uranium Enrichment (wt.% 235U)

Plutonium Content (wt.% 239Pu)

Atom ratio of hydrogen to fissile material (H/X)

Each parameter was plotted against the knorm results independently for each case that was analyzed.

These plots are provided in Figure 9 through Figure 12. This scatter plot of data was first analyzed by visual inspection to determine if any trends were readily apparent in the data. During this inspection, the axes of the graphs were modified to different scales to allow for a more thorough review. No clear evidence of a trend, linear or otherwise, was observed from this inspection.

((

` ` ` ` ` ))

Figure 9 - Scatterplot of knorm versus EALF 41

NEDO-33932 Revision 1 Non-Proprietary Information

((

` ` ` ` ` ))

Figure 10 - Scatterplot of knorm versus 235U wt.%{3}

((

` ` ` ` ` ))`

Figure 11 - Scatterplot of knorm versus 239Pu wt.%

42

NEDO-33932 Revision 1 Non-Proprietary Information

((

` ` ` ` ` ))

Figure 12 - Scatterplot of knorm versus H/X To further check for trends in the data, a linear regression was performed. The linear regression fitted equation is in the form y(x)= a +bx, where y is the dependent variable (knorm) and x is any of the predictor variables from Table 18. Unweighted knorm values were used in this evaluation, although it is noted that, due to the very similar values reported in Table 17, using weighted values would produce very similar results. This regression was performed using the built-in regression analysis tool in Excel. The fitted lines are included in Figure 9 through Figure 12.

Again, it is noted through visual inspection that the trends do not appear to exhibit a strong correlation to the data. A useful tool to validate this claim is the linear correlation coefficient.

This is a quantitative measure of the degree to which a linear relation exists between two variables.

It is often expressed as the square term, r2, and can be calculated directly using built in functions in Excel. The closer r2 gets to the value of 1, the better the fit of data is expected to be to the linear equation. Results from this linear regression evaluation are summarized in Table 19.

A final method to test for goodness of fit is the chi squared test (2). This method is explained in detail in Reference 6. In general, it can be stated that 2 is an indicator of the agreement between the observed (calculated) and expected (fitted) values for some variable. For linear goodness of fit testing using this method, Equation A-3 is utilized, where the expected value of f(xi) corresponds to the linear fitted equation for the trending parameter, xi.

= (A-3) 43

NEDO-33932 Revision 1 Non-Proprietary Information A more convenient way to report this result is the reduced chi squared value, which is denoted as and is defined by Equation A-4, where d is the degrees of freedom for the evaluation.

= / (A-4)

If a value of one or less is obtained for this equation, then there is no reason to doubt the expected (fitted) distribution is reasonable; however, if the value is much larger than one, the expected distribution is unlikely to be a good fit. Results for each trending parameter are summarized in Table 19.

Table 19 - Trending Results Summary Trend Valid Intercept Slope r2 Parameter Trend EALF ((` ` ` ` ` ` ` `````````` ``````` ``` No 235 U wt.% ``````` `````````` ``````` ``` No 239 Pu wt.% ``````` ``````````` ``````` ``` No H/X ``````` ``````````` ``````` ` ` ` ` ` ` ` )) No The results in Table 19 clearly demonstrate that there are no statistically significant or valid trends of knorm with any of the trending parameters.

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 13. 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.

44

NEDO-33932 Revision 1 Non-Proprietary Information

((

` ` ` ` ` ))

Figure 13 - 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) n knorm i i =1 t2 k norm = n 1 (A-7) i =1 2

t (A-8)

SP = s2 + 2 n

2 = n (A-9) 1 i =1 2

t 45

NEDO-33932 Revision 1 Non-Proprietary Information 2

1 n 1 2 (k norm i k norm )

2 n 1 i =1 t s = (A-10) 1 n 1 n i =1 t2 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 20 summarizes the results of these calculations.

Table 20 - 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 ` ` ` ` ` ` ` ` ` ` ))

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/CR-6698 (Reference 5) guidance. As seen in Figure 13, ((` ` ` ` ` ` ` ` ` ` ` ` `

````````````````````````````````````````````````````````````````````````````````````````````````

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `` ` ` ` )) As shown in Table 20, the MCNP-05P bias uncertainty (95/95) ((` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` `

` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` )) Table 21 summarizes the recommended bias and bias uncertainty to be used in criticality calculations.

Table 21 - Recommended Bias and Bias Uncertainty in Criticality Analyses for MCNP-05P with ENDF/B-VII Bias ((` ` `

Bias Uncertainty (95/95) ` ` ` ` ` ` ` ` ` ` ` ))

46

NEDO-33932 Revision 1 Non-Proprietary Information APPENDIX B - LEGACY FUEL STORAGE JUSTIFICATION Exposure dependent, maximum, uncontrolled in-core k results for each fuel assembly in the Quad Cities spent fuel pools are confirmed to be less than 1.29. The in-core k values have been calculated using the process for validating that specific assembly designs are acceptable for storage in the Quad Cities fuel storage racks, as outlined in Section 3.5, and the in-core reactivity values are presented in Table 22. This information demonstrates that all fuel assemblies currently in the Quad Cities spent fuel pool have considerable margin to the reactivity of the GNF3 design basis bundle used in this analysis. Any GNF3 bundles in the Quad Cities core or spent fuel pool are covered by the design basis bundle study in Section 5.3.

The GNF3 design basis bundle with an in-core k value of 1.29 was shown to be below the 10 CFR 50.68 0.95 in-rack k-effective limit when analyzed in the storage racks. As represented in Table 22, the limiting legacy GNF fuel type and limiting legacy non-GNF fuel provided by Exelon have a significantly lower in-core k value than the GNF3 design basis bundle (i.e., less reactive than the design basis bundle). Therefore, it is confirmed that all legacy fuel bundles are safe for storage in the Quad Cities spent fuel storage racks with rack inserts installed.

Table 22 - Limiting Cold As-Designed Eigenvalue of Bundles Inserted Into Quad Cities Bundle Bundle Name In-Core k

((` ` ` ` ` ` ` `` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ` ``````

```` ``````````````````````````````````` ` ` ` ` ` ` ` ` ` ` ))

47

NEDO-33932 Revision 1 Non-Proprietary Information Table 23 shows a list of legacy fuel bundles in Quad Cities spent fuel pools that have empty rod locations. The in-core k values have been calculated with all rods in their original locations using the methodology outlined in Section 3.5. The margin to safety was confirmed to exist in the storage racks by analyzing the bundles with the corresponding rods removed to reflect their current state. These bundles were analyzed under nominal storage conditions, as outlined in Section 5.4, and the in-rack reactivity values are presented in Table 23. The GNF3 design basis bundle, with a nominal in-rack k value reported in Table 10, was shown to be below the 10 CFR 50.68 kmax(95/95) limit of 0.95 when analyzed in the storage racks. As represented in Table 23, the legacy bundles with missing rod locations have a significantly lower in-rack k value than the GNF3 design basis bundle (i.e., less reactive than the design basis bundle). Therefore, it is confirmed that these legacy fuel bundles with missing rod locations are safe for storage in the Quad Cities spent fuel storage racks with rack inserts installed.

Table 23 - Legacy Bundles with Missing Rod Locations at Quad Cities In-Rack

Rods Assembly Bundle Name & Description Nominal Missing Reactivity

((` ` ` ` ` ` ` ```````

```````````````````````````````````````````

```````````

`````` ` ```````

`````````````````````````````````

```````````

`````` ` ```````

`````````````````````````````````

``````````

`````` ` ```````

```````````````````````````````````

``````````

`````` ` ```````

```````````````````````````````````

`````` ` ` ` ` ` ` ` ` ` ` ` ` ))

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48

ATTACHMENT 4 NEI 12-16 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-33932P and Section 1.0 of Attachment 1 Changes requested YES Section 1.0 of NEDC-33932P and Sections 2.1, 2.2, and 2.3 of Attachment 1 Summary of physical changes YES Section 1.0 of NEDC-33932P and Sections 2.1, 2.2, and 2.3 of Attachment 1 Summary of Tech Spec changes YES Section 2.3 of Attachment 1 and Attachment 2 Summary of analytical scope YES Sections 1.0 and 3.0 of NEDC-33932P and Sections 2.1 and 2.2 of Attachment 1 2.0 Acceptance Criteria and Regulatory Guidance Summary of requirements and guidance YES Section 2.0 of NEDC-33932P and Sections 2.1 and 2.2 of Attachment 1 Requirements documents referenced YES Section 2.0 of NEDC-33932P and Sections 2.1 and 2.2 of Attachment 1 Guidance documents referenced YES Section 2.0 of NEDC-33932P and Sections 2.1 and 2.2 of Attachment 1 Acceptance criteria described YES Section 2.0 of NEDC-33932P and Sections 2.1 and 2.2 of Attachment 1 3.0 Reactor and Fuel Design Description Describe reactor operating parameters NO Not applicable for this analysis. See Sections 3.7 and 5.5 of NEDC-33932P for depletion parameters and assumptions.

Describe all fuel in pool YES Section 4.0 of NEDC-33932P C-1

Subject Included Notes / Explanation Geometric dimensions (Nominal and YES Section 4.1 of NEDC-33932P Tolerances)

Schematic of guide tube patterns NO Not applicable for BWR fuel Material compositions YES Section 4.0 of NEDC-33932P Describe future fuel to be covered YES Section 4.0 of NEDC-33932P Geometric dimensions (Nominal and YES Section 4.1 of NEDC-33932P Tolerances)

Schematic of guide tube patterns NO Not applicable for BWR fuel Material compositions YES Section 4.0 of NEDC-33932P Describe all fuel inserts NO There are no fuel inserts in analysis Geometric Dimensions (Nominal and NEDC-33932P.

Tolerances)

Schematic (axial/cross-section)

Material compositions Describe non-standard fuel YES Section 4.0, Appendix B of Geometric dimensions NEDC-33932P Describe non-fuel items in fuel cells YES Section 4.0 of NEDC-33932P Nominal and tolerance dimensions NO Not applicable; analysis NEDC-33932P covers peak reactivity in every 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-33932P. See Section 2.2 of Attachment 1 for details.

Spent fuel pool, Storage rack description YES Sections 5.1-5.2 of NEDC-33932P and Nominal and tolerance dimensions Section 3.1 of Attachment 1 Schematic (axial/cross-section)

Material compositions Other Reactivity Control Devices (Inserts) YES Sections 5.1-5.2 of NEDC-33932P and Nominal and tolerance dimensions Section 3.1 of Attachment 1 Schematic (axial/cross-section)

Material compositions 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-33932P. See Section 2.2 of Accident analysis Attachment 1 for details.

Spent fuel storage rack analysis description YES Sections 3.5-3.7 and 5.0 of NEDC-33932P Storage geometries YES Sections 5.1-5.2 of NEDC-33932P C-2

Subject Included Notes / Explanation Bounding assembly design(s) YES Section 5.3 of NEDC-33932P Soluble boron credit NO Not applicable - No soluble boron credit in this BWR criticality analysis Boron dilution analysis (NEDC-33932P)

Burnup credit NO No burnup credit in BWR peak reactivity analysis NEDC-33932P -

fuel is evaluated at peak reactivity Decay/Cooling time credit NO No decay/cooling time credit in analysis of NEDC-33932P Integral absorber credit YES Sections 5.1-5.2 of NEDC-33932P Other credit NO No other credit in analysis NEDC-33932P Fixed neutron absorbers YES Rack inserts - Sections 5.1-5.2 of NEDC-33932P Aging management program NO Aging is not included in analysis NEDC-33932P; no credit is taken for Boraflex Accident analysis YES Section 5.5.3 of NEDC-33932P Temperature increase YES Sections 5.4-5.5 of NEDC-33932P Assembly drop YES Section 5.5.3 of NEDC-33932P Single assembly misload YES Section 5.5.3 of NEDC-33932P Multiple misload NO Uniform pool, no opportunity for multiple misload Boron dilution NO Not applicable - No soluble boron credit in this BWR criticality analysis (NEDC-33932P)

Other YES Sections 5.5.3 of NEDC-33932P Fuel out of rack analysis YES Section 5.5 of NEDC-33932P considers Handling worst case abnormal positioning of a Movement fuel assembly outside the storage rack.

Inspection 6.0 Computer Codes, Cross Sections and Validation Overview Code/Modules Used for Calculation of keff YES Section 3.0 of NEDC-33932P Cross section library YES Section 3.1 of NEDC-33932P Description of nuclides used YES Section 4.2 of NEDC-33932P Convergence checks YES Section 3.3 of NEDC-33932P Code/Module Used for Depletion YES Section 3.0 of NEDC-33932P Calculation Cross section library YES Section 3.1 of NEDC-33932P Description of nuclides used YES Section 4.2 of NEDC-33932P Convergence checks YES Section 3.3 of NEDC-33932P Validation of Code and Library YES Section 3.4, Appendix A of NEDC-33932P C-3

Subject Included Notes / Explanation Major Actinides and Structural YES Section 3.4 of NEDC-33932P Materials Minor Actinides and Fission Products YES Section 3.4 of NEDC-33932P Absorbers Credited YES Section 3.4 of NEDC-33932P 7.0 Criticality Safety Analysis of the New Fuel Rack Rack model YES The NFV rack CSA coverage for the Boundary conditions new GNF3 fuel will be the GESTAR II Source distribution analysis for the GE designed low Geometry restrictions density NFV racks upon approval of Limiting fuel design this license amendment. The QCNPS Fuel density NFV racks are GE designed low Burnable Poisons density racks with an interrack spacing Fuel dimensions of 11 inches, which is 10.5 inches Axial blankets (the criteria listed in GESTAR II) and Limiting rack model thus, the racks may be utilized to store Storage vault dimensions and materials new GNF fuel with in-core SCCG Temperature kinf 1.31. See Section 2.2 of Attachment 1 for details.

Multiple regions/configurations 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 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 C-4

Subject Included Notes / Explanation 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-33932P Convergence verification Simplifications Non-uniform enrichments 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-33932P Boundary conditions Source distribution Geometry restrictions Design Basis Fuel Description YES Section 5.3 of NEDC-33932P Fuel density YES Section 4.1 of NEDC-33932P Burnable Poisons YES Section 5.2 of NEDC-33932P Fuel assembly inserts NO No fuel assembly inserts in analysis of NEDC-33932P Fuel dimensions YES Section 4.1 of NEDC-33932P Axial blankets NO Section 3.7 of NEDC-33932P Configurations considered YES Section 6.0 of NEDC-33932P Borated NO Not applicable for this BWR analysis (NEDC-33932P)

Unborated YES BWR analysis NEDC-33932P considers unborated SFP Multiple rack designs NO Only one spent fuel rack design is present at Quad Cities and the NFV racks are covered by the GESTAR II analysis (see Section 2.2 of Attachment 1).

Alternate storage geometry NO Not applicable for analysis of NEDC-33932P Reactivity Control Devices YES Sections 5.1- 5.2 of NEDC-33932P Fuel Assembly Inserts NO No fuel assembly inserts in analysis NEDC-33932P C-5

Subject Included Notes / Explanation Storage Cell Inserts YES Sections 5.1- 5.2 of NEDC-33932P Storage Cell Blocking Devices NO No blocking devices in analysis NEDC-33932P Axial burnup shapes NO Section 3.7 of NEDC-33932P Uniform/Distributed YES Section 3.7 of NEDC-33932P Nodalization NO Section 3.7 of NEDC-33932P Blankets modeled NO Section 3.7 of NEDC-33932P Tolerances/Uncertainties YES Section 5.6 of NEDC-33932P Fuel geometry Fuel rod pin pitch Fuel pellet OD Cladding OD Axial fuel position NO Section 3.7 of NEDC-33932P Fuel content YES Section 5.6 of NEDC-33932P Enrichment Density Assembly insert dimensions and NO No fuel assembly inserts in analysis materials NEDC-33932P Rack geometry YES Section 5.6 of NEDC-33932P Flux-trap size (width) NO Not applicable to non-flux-trap racks Rack cell pitch YES Section 5.6 of NEDC-33932P Rack wall thickness YES Section 5.6 of NEDC-33932P Neutron Absorber Dimensions YES Section 5.6 of NEDC-33932P Rack insert dimensions and materials YES Section 5.6 of NEDC-33932P Code validation uncertainty YES Sections 3.4, 5.6, and Appendix A of NEDC-33932P Criticality case uncertainty YES Section 5.6 of NEDC-33932P Depletion Uncertainty YES Sections 3.4, 5.6 of NEDC-33932P Burnup Uncertainty NO Not applicable for BWR peak reactivity analysis NEDC-33932P Biases YES Section 5.0 of NEDC-33932P Design Basis Fuel design YES Section 5.3 of NEDC-33932P Code bias YES Sections 3.4, 5.5 of NEDC-33932P Temperature YES Section 5.4 of NEDC-33932P Eccentric fuel placement YES Sections 5.4-5.5 of NEDC-33932P Incore thimble depletion effect NO Not applicable for analysis NEDC-33932P NRC administrative margin NO Not applicable for analysis NEDC-33932P Modeling simplifications YES Sections 3.7, 4.2 of NEDC-33932P Identified and described C-6

Subject Included Notes / Explanation 10.0 Interface Analysis Interface configurations analyzed NO N/A, the spent fuel pool is uniform Between dissimilar racks NO with rack inserts in every cell. Only Between storage configurations within NO one rack design.

a rack Interface restrictions NO 11.0 Normal Conditions Fuel handling equipment NO Not in the scope and does not impact results of criticality analysis NEDC-33932P.

Administrative controls NO No new administrative controls included in NEDC-33932P or Attachment 1.

Fuel inspection equipment or processes NO Not in the scope and does not impact results of criticality analysis NEDC-33932P.

Fuel reconstitution YES Section 4.0 of NEDC-33932P 12.0 Accident Analysis Boron dilution NO Not applicable - No soluble boron Normal conditions credit in this BWR criticality analysis Accident conditions (NEDC-33932P)

Single assembly misload YES Section 5.5 of NEDC-33932P Fuel assembly misplacement YES Section 5.5 of NEDC-33932P Neutron Absorber Insert Misload YES Section 5.5 of NEDC-33932P Multiple fuel misloads NO Uniform pool, single storage configuration, no opportunity for multiple misloads Dropped assembly YES Section 5.5 of NEDC-33932P Temperature YES Section 5.4 of NEDC-33932P Seismic event/other natural phenomena YES Section 5.5 of NEDC-33932P 13.0 Analysis Results and Conclusions Summary of results YES Section 6.0 of NEDC-33932P NO Not applicable for BWR peak reactivity Burnup curve(s) analyses, including NEDC-33932P NO Not applicable for BWR peak reactivity Intermediate Decay time treatment analyses, including NEDC-33932P NO No new administrative controls New administrative controls included in NEDC-33932P or Attachment 1.

YES Section 2.3 of Attachment 1 and Technical Specification markups Attachment 2 C-7

Subject Included Notes / Explanation 14.0 References YES Section 7.0 of NEDC-33932P Appendix A: Computer Code Validation: Appendix A of NEDC-33932P Code validation methodology and bases YES Appendix A of NEDC-33932P New Fuel Depleted Fuel MOX HTC Convergence Trends Bias and uncertainty Range of applicability YES Described in Section 3.4 of NEDC-33932P Analysis of Area of Applicability YES Described in Section 3.4 of coverage NEDC-33932P 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-33932P, Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis, Revision 1, October 2021. 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-33932P Revision 1 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-33932P Revision 1 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 15th day of October 2021.

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-33932P Revision 1 Affidavit Page 3 of 3

ATTACHMENT 6 Curtiss-Wright Corporation 10 CFR 2.390 Affidavit

AFFIDAVIT I, Matthew C. Harris, Segment Manager of NETCO, business segment of Scientech, Curtiss-Wright Corporation, do hereby affirm and state:

1. I am the Segment Manager of the NETCO, business segment of Scientech, Curtiss-Wright Corporation, and am authorized to execute this affidavit on its behalf. I am further authorized to review information submitted to the Nuclear Regulatory Commission (NRC) and apply to the NRC for the withholding of information from disclosure.

2. The information sought to be withheld is contained in the GNF Report Quad Cities Units 1 and 2: Fuel Storage Criticality Safety Analysis - NEDC-33932P, Revision 1, October 2021. Curtiss-Wright Flow Control Service, Co. LLC confidential proprietary information is identified by a solid underline inside double square brackets. ((This sentence is an example. {C})) Curtiss-Wright proprietary information in Figures and large objects is identifiable by double square brackets before and after the object.

3. In making this application for withholding of proprietary information of which it is the owner, NETCO relies on provisions of NRC regulation 10 CFR 2.390(a)(4). The information for which exemption from disclosure is sought is confidential commercial information.

4. The proprietary information provided by NETCO should be held in confidence by the NRC pursuant to the policy reflected in 10 CFR 2.390(a)(4) because:

a) The information sought to be withheld in the Report (see paragraph 2 above) is and has been held in confidence by NETCO.

b) This information is of a type that is customarily held in confidence by NETCO, and there is a rational basis for doing so because the information contains methodology, data and supporting information developed by NETCO that could be used by a competitor as a competitive advantage.

c) This information is being transmitted to the NRC in confidence.

Page 1 of 2 Curtiss Wright Nuclear Division 44 Shelter Rock Rd

Danbury, CT 06810

Phone: 203.448.3439

Fax: 203.437.6279

d) This information sought to be withheld, to the best of my knowledge and belief, is not available in public sources and no public disclosure has been made.

e) The information sought to be withheld contains developed, patented, product fabrication data and supporting information that could be used by a competitor as a competitive advantage, and would result in substantial harm to the competitive position of NETCO. This information would reduce the expenditure of resources and improve his competitive position in the implementation of a similar product. Third party agreements have been established to ensure maintenance of the information in confidence. The development of the methodology, data and supporting information was achieved at a significant cost to NETCO. Public disclosure of this information sought to be withheld is likely to cause substantial harm to NETCO's competitive position and reduce the availability of profit-making opportunities.

5. Initial approval of proprietary treatment of a document is made by the Segment Manager of NETCO, business segment of Scientech, the person most likely to be familiar with the value and sensitivity of the information and its relation to industry knowledge. Access to such information within NETCO is on a "need to know" basis.

6. Accordingly, NETCO requests that the designated document be withheld from public disclosure pursuant to 10 CFR 2.390(a)(4).

I declare under penalty of perjury that the foregoing affidavit and statements therein are true and correct to the best of my knowledge, information and belief.

Matthew C. Harris Segment Manager, NETCO , business segment of Scientech Curtiss-Wright Corporation Date: __10/15/21__________

Page 2 of 2 Curtiss Wright Nuclear Division 44 Shelter Rock Rd

Danbury, CT 06810

Phone: 203.448.3439

Fax: 203.437.6279

v t e Letter Sequence Request
EPID:L-2021-LLA-0196, Licensee Outline Submittal Email (Quad Cities, 2023) (Approved, Closed)
Initiation Request, Request Acceptance Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Questions 13 June 2022 Answers 13 July 2022 Results Approval Other:
MONTHYEAR2022-12-02 [Table View]

RS-21-065, 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

Text

Subject:

Reference:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

1.0 INTRODUCTION

6.0 CONCLUSION

7.0 REFERENCES

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