View source for RS-22-090, Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies

{{Adams
| number = ML22194A085
| issue date = 07/13/2022
| title = Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies
| author name = Simpson P
| author affiliation = Constellation Energy Company, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000254, 05000265
| license number = DPR-029, DPR-030
| contact person = 
| case reference number = RS-22-090
| document report number = NEDO-33932, Rev 2
| package number = ML22194A084
| document type = Legal-Affidavit, Letter type:RS, Report, Technical, Response to Request for Additional Information (RAI)
| page count = 79
| project = 
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 July 13, 2022                                                                           10 CFR 50.90 RS-22-090 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 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:==
Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies

==References:==
: 1. Letter from P.R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "
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," dated October 25, 2021 (ADAMS Accession No. ML21298A168)
: 2. Email from R. Kuntz (U.S. NRC) to R. Steinman (Constellation Energy Generation), "RAI RE: Quad Cities amendment for spent fuel pool storage analysis (EPID L-2021-LLA-0196)," dated June 13, 2022 (ADAMS Accession No. ML22164A785)
In Reference 1, Constellation Energy Generation, LLC (CEG) requested 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. The proposed changes support the transition from Framatome (formerly AREVA) ATRIUM 10 XM fuel to Global Nuclear Fuel - Americas, LLC (GNF-A) GNF3 fuel at QCNPS by allowing a different methodology to be used for the criticality safety evaluation for the spent fuel pool (SFP) and the new fuel vault (NFV).
In Reference 2, the NRC requested additional information that is needed to complete review of the proposed methodology change. Attachments 1 (non-proprietary) and 6 (proprietary) provide the additional information requested. A signed affidavit from the owner of the information, GNF-A, is included as Attachment 3. The affidavit sets forth the basis on which GNF-As 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 in Attachments 6 and 7 contain Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachments 6 and 7, this document is decontrolled.

U.S. Nuclear Regulatory Commission July 13, 2022 Page 2 , which is proprietary to GNF-A be withheld from public disclosure.
Additionally, a typographical error was identified in Section 5.5.3, "Abnormal / Accident Bias Cases," of the versions of NEDO-33932 / NEDC-33932P, "Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis" that were originally submitted as Attachments 3 and 5 of Reference 1, respectively. The attached revised copies of the non-proprietary (Attachment 2) and proprietary (Attachment 7) versions of the report replace the previously submitted versions in their entirely, although only page 26 of the document is changed. contains information proprietary to GNF-A and Curtiss-Wright. As a result, this document is supported by signed affidavits from the owners of the information, which are included as Attachments 4 and 5, 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 in Attachment 7, which is proprietary to GNF-A and Curtiss-Wright be withheld from public disclosure.
CEG has reviewed the information supporting the finding of no significant hazards consideration, and the environmental consideration that were previously provided to the NRC in Reference 1. The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. In addition, the information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.
CEG is notifying the State of Illinois of this supplement to a previous application for a change to the operating license by sending a copy of this letter and its attachments to the designated State Official in accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"
paragraph (b).
There are no regulatory commitments included in this letter.
Should you have any questions concerning this letter, please contact Ms. Rebecca L. Steinman at 630-657-2831.

U.S. Nuclear Regulatory Commission July 13, 2022 Page 3 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 13th day of July 2022.
Respectfully, Patrick R. Simpson Sr. Manager Licensing Constellation Energy Generation, LLC Attachments:
: 1. Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies (Non-Proprietary)
: 2. NEDO-33932, "Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis,"
Revision 2, dated July 2022 (Non-Proprietary)
: 3. Global Nuclear Fuels - Americas, LLC 10 CFR 2.390 Affidavit for Attachment 6
: 4. Global Nuclear Fuels - Americas, LLC 10 CFR 2.390 Affidavit for Attachment 7
: 5. Curtiss-Wright Corporation 10 CFR 2.390 Affidavit for Attachment 7
: 6. Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies (Proprietary)
: 7. NEDC-33932P, "Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis,"
Revision 2, dated July 2022 (Proprietary) cc:     Regional Administrator - NRC Region III NRC Senior Resident Inspector - Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Department of Nuclear Safety

ATTACHMENT 1 QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 Docket Nos. 50-254 and 50-265 Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies (Non-Proprietary)

ATTACHMENT 1 Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies (Non-Proprietary Version)
Regulatory Requirements Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix A, Criterion 5 requires, Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown.
10 CFR Part 50, Appendix A, Criterion 62 requires, Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
Paragraph 50.68(a) of 10 CFR requires, Each holder of a construction permit or operating license for a nuclear power reactor issued under this part or a combined license for a nuclear power reactor issued under Part 52 of this chapter, shall comply with either 10 CFR 70.24 of this chapter or the requirements in paragraph (b) of this section. The licensee has chosen to comply with Paragraph 50.68(b) of 10 CFR.
Paragraph 50.68(b)(1) of 10 CFR requires, Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.
Paragraph 50.68(b)(2) of 10 CFR requires, The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.
Paragraph 50.68(b)(3) of 10 CFR requires, If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.
Paragraph 50.68(b)(4) of 10 CFR requires, in part, If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95-percent probability, 95-percent confidence level, if flooded with unborated water.
The QCNPS SFP NCS analysis does not contain soluble boron, so the 10 CFR 50.68(b)(4) requirements regarding soluble boron do not apply.
In addition, paragraph 50.36(c)(4) of 10 CFR requires, Design features. Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c) (1), (2), and (3) of this section.
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ATTACHMENT 1 Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies (Non-Proprietary Version)
RAI-SFNB-1 In Section 2.2 New Fuel Vault Criticality Safety Analysis of Attachment 1 to the {{letter dated|date=October 25, 2021|text=October 25, 2021, letter}} it states, 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 [Standard Cold Core Geometry] 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.
However, neither Reference 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 (ML20330A199) nor Reference 6.6 NEDC-33879P, Revision 4, GNF3 Generic Compliance with NEDE-24011-P-A (GESTAR II)," dated August 2020 (ML20244A110) contain a nuclear criticality safety methodology or nuclear criticality safety analysis. To evaluate compliance with 10 CFR 50.68(b)(2) submit the following:
* NFV criticality safety analysis methodology used in the analysis.
* Criticality safety analysis that sets the limits for the QCNPS NFV.
* Criticality safety analysis that demonstrates GNF3 meets the limits for the QCNPS NFV.
CEG Response to RAI-SFNB-1 The New Fuel Vault (NFV) criticality safety analysis methodology - the peak, cold in-core lattice infinite multiplication factor (kinf) criterion for demonstrating compliance to the 10 CFR 50.68 fuel storage criticality criterion - has been previously approved as part of GESTAR II. Revision 31 to NEDE-24011P - GESTAR II is the latest version of GESTAR II. The use of the NFV methodology with GNF3 fuel was previously reviewed by the U.S Nuclear Regulatory Commission (U.S. NRC) as part of Amendment 37 to GESTAR II, which was incorporated into NEDE-24011P at Revision 24. Amendment 37 RAI-3 (Reference 2), which is included on page US B-181 of the U.S. Supplement to GESTAR II (Reference 3), requested clarification regarding the applicability of the kinf calculations for the lattices supporting the NFV analysis to the current GNF fuel products. The response to Amendment 37 RAI-3 provided methodology details and cited other NRC approved reports utilizing the same methodology. The peak, cold in-core lattice infinite multiplication factor (kinf) criterion for demonstrating compliance to the 10 CFR 50.68 fuel storage criticality criterion has been used for all GE supplied fuel storage racks and is currently used for re-rack designs at a number of plants. The methodology relies upon a well-characterized linear relationship between in-core kinf and in-rack keff, which is evaluated for each rack. A conservative lattice with a peak, cold in-core kinf value at or above the intended storage limit is used in the criticality analyses. A criticality analysis is performed for each new GNF fuel product line per GESTAR II Section 1.1.3.G, which confirms that the kinf limits described in GESTAR II Section 3.5 would result in a kmax value compliant with 10 CFR 50.68. The issued Safety Evaluation (SE) for Amendment 37 to GESTAR II concluded 2 of 8

ATTACHMENT 1 Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies (Non-Proprietary Version) that the criteria and methodology used to evaluate fresh and irradiated fuel were acceptable (Reference 4).
GNF performed a criticality safety analysis for GNF3 fuel using bounding new fuel storage rack parameters in support of their GNF3 GESTAR II validation report. The analysis performed is generic to all NFV racks manufactured by GE, including those installed at QCNPS. The generic NFV criticality analysis modeled the actual rectangular dimensions and tolerances of two rack options to determine the restrictions outlined in GESTAR II. The two models used in the generic GNF3 criticality safety analysis have rectangular dimensions of ((               )) x 10.5 inches and ((                            )). The limiting model bounds the storage geometry of the NFV racks at QCNPS. The applicable generic criticality analysis performed for GNF3 fuel is based on the geometry of the NFV racks in which it is to be stored, specifically racks with an interrack pitch  10.5 inches. Per GESTAR II, NFV racks with an interrack pitch  10.5 inches may be utilized to store new GNF fuel with an in-core SCCG kinf  1.31. Since the QCNPS interrack pitch meet this restriction, the QCNPS NFV racks may be utilized to store new GNF3 fuel with in-core SCCG kinf  1.31.
RAI-SFNB-2 QCNPS Updated Final Safety Analysis Report (UFSAR) section 9.1.1.2 states, The minimum center-to-center spacing of bundles in the racks is 6.625 inches longitudinally by 11 inches between rows. This is not a pitch  10.5 inches. To evaluate compliance with 10 CFR 50.68(b)(2) explain and clarify this apparent discrepancy.
CEG Response to RAI-SFNB-2 The referenced UFSAR statement means that within a row the bundles are placed 6.625 inches apart (center-to-center), and from one row to the next the bundles are placed 11 inches apart (center-to-center). The geometry of the NFV racks has been confirmed from plant drawings to be 6.625 inches along a row and 11 inches from row to row. This nominal spacing has not changed since construction of the NFV racks. As stated in the response to RAI-SFNB-3, the mechanical and structural design of the NFV racks ensures that this nominal spacing between bundles is maintained at all times, even during the worst-case seismic event.
The interrack spacing referred to in GESTAR II is the spacing between rows. For QCNPS, this spacing is 11 inches, which is greater than 10.50 inches required by GESTAR II. See the response to RAI-SNFB-1 for the details regarding the basis for the 10.5 inch limit.
RAI-SFNB-3 The QCNPS NFV center to center pitch is critical to maintaining the geometric spacing of fuel assemblies to ensure 10 CFR 50.68(b)(2) is met. Describe the controls QCNPS has in place to ensure the QCNPS NFV center to center spacing is maintained.
CEG Response to RAI-SFNB-3 The mechanical and structural design of the NFV racks ensures that the spacing between bundles in the NFV racks is maintained at all times, even during the worst-case seismic event.
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ATTACHMENT 1 Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies (Non-Proprietary Version)
This robustness is captured in the following QCNPS UFSAR statements:
* Section 9.1.1.1 which states The new fuel storage vault is designed to withstand earthquake loading as a Class I structure.
* Section 9.1.1.2, This new fuel storage vault is a reinforced concrete Class I structure, accessible only through top hatches.
In addition, the physical design of the NFV racks prevents bundles from being inserted into locations within the racks that would reduce the physical spacing such that the generic criticality safety analysis for new GNF3 would no longer be applicable.
Therefore, the spacing in the NFV racks is ensured to be greater than that required in the GESTAR II criticality safety analysis for new GNF3 fuel stored in an NFV.
RAI-SFNB-4 In Section 4.1 Applicable Regulatory Requirements/Criteria of Attachment 1 to the {{letter dated|date=October 25, 2021|text=October 25, 2021, letter}} states, in part 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). 10 CFR 50.68(b)(3) requires optimum moderation be prevented to forgo complying with the k effective portion of the paragraph. Section 9.1.1.3 of the QCNPS UFSAR does not list any means by which an optimum moderation condition is prevented. To evaluate compliance with 10 CFR 50.68(b)(3) explain/justify how an optimum moderation condition is precluded at all times.
CEG Response to RAI-SFNB-4 SIL 152, "Criticality Margins for Storage of New Fuel," (Reference 9) identified potential sources for an optimum moderator such as fire extinguisher foam, water mist, steam or other hydrogenous materials and recommended procedural guidance for new fuel handling operations.
As described in Section 9.1.1.3 of the QCNPS UFSAR, QCNPS has identified the Refueling Floor (which includes the NFV storage array) as a moderation controlled area and adopted administrative controls aligned with SIL 152 guidance to limit the amount of hydrogenous material to preclude the existence of the optimum moderation condition in the new fuel vault.
Specific examples of the administrative and procedural controls adopted at QCNPS include:
 -    The Refueling Floor does not have fog-type firefighting nozzles. Only smooth bore nozzles are provided.
 -    Only manual firefighting equipment is utilized on the Refueling Floor and adequate floor drainage is provided.
 -    New fuel should NOT be stored in the NFV when there are construction activities on the refueling floor OR construction debris in the vicinity of the NFV unless a solid, fireproof cover is placed over the vault to preclude criticality due to inundation by low density water such as water fog or spray from a fire hose.
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ATTACHMENT 1 Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies (Non-Proprietary Version)
  -  Load handling above the fuel storage area (NFV with plug(s) removed) is to be limited to one fuel assembly or weight equivalent.
  -  A fuel array of up to three fuel bundles outside of a normal storage area or normal shipping container should be maintained with an edge-to-edge spacing of 12 inches or more from all other fuel.
  -  A fuel array of four or more fuel bundles outside of the normal fuel storage areas or normal shipping containers is prohibited.
  -  No more than three (3) new fuel bundles/assemblies may be in an interim storage location at any one time (approved storage locations are shipping containers, the NFV and the Spent Fuel Storage Pool).
  -  Fuel handling in the fuel storage area should be limited to the normal weight limit per hoist.
Exceptions to this requirement are interlock test weights and fuel pool gates. Other equipment may be moved over the fuel storage area if necessary to do so, provided a specific procedure has been written and approved prior to the move and that travel over the area is kept to the minimum extent necessary to make the move.
  -  The NFV should always be kept dry.
  -  Fuel movement in the NFV must not be permitted if an abnormal condition of vault flooding occurs.
  -  Fuel should NOT be placed in aisles or moved through aisles adjacent to and at the same level of the storage racks.
RAI-SFNB-5 The measures QCNPS has to ensure NFV optimum moderation condition is precluded at all times are essential to forgoing the NFV optimum moderation k effective analysis otherwise stipulated in 10 CFR 50.68(b)(3). Describe the controls QCNPS has in place to ensure those measures are not compromised.
CEG Response to RAI-SFNB-5 The measures that preclude the NFV optimum moderation condition, as discussed in the CEG response to RAI-SFNB-4, are ensured to not be compromised through use of controls that are implemented via the Fire Protection Program or other procedural requirements. This is summarized as follows:
* Hydrogenous material is limited on the refuel floor at all times.
* The QCNPS Fire Protection Report (FPR) dictates the fire plan for the refueling floor including which firefighting equipment is to be used or not used.
* The fuel handling guidance described in RAI-SFNB-4 is specified in QCNPS fuel handling procedures.
As allowed by 10 CFR 50.68(b)(3), analyzing the optimum moderation condition is not required since administrative controls and/or design features are in place to prevent such moderation.
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ATTACHMENT 1 Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies (Non-Proprietary Version)
RAI-SFNB-6 In Section 1.0, INTRODUCTION" of NEDC-33932P Revision 1 (Attachment 3 to the {{letter dated|date=October 25, 2021|text=October 25, 2021, letter}}) it states, A maximum SCCG, uncontrolled peak in-core k
[eiqenvalue] of 1.29 as defined by the lattice physics code TGBLA06 (Reference 1) is set as the limit for this analysis. However, NEDC-33932P Revision 1 Reference 1 does not have a clear nexus to how TGBLA06 calculates a SCCG. Additionally, NEDC-33932P Revision 1 Reference 1 is dated November 10, 1999, which predates GNF3 fuel by at least a decade. To evaluate compliance with 10 CFR 50.68(b)(4) provide the methodology or appropriate reference for how TGBLA06 calculates SCCG and the analysis or appropriate reference for how TGBLA06 is an appropriate code for modeling GNF3.
CEG Response to RAI-SFNB-6 TGBLA06 is a lattice physics code that calculates the exposure dependent pin-by-pin isotopic specifications used in developing the design basis lattice for the spent fuel pool criticality safety analysis, but also has application to many other GEH/GNF analysis methods. Reference 1 of NEDC-33931P Revision 1 documents the NRCs original acceptance of TGBLA06 and its associated application methodology (Reference 8). This initial approval was updated to address the applicability to the GNF3 product line in Amendment 49 to NEDE-24011P - GESTAR II (Reference 5), Supplement 5P-A, Revision 1, Applicability of GE Methods to Expanded Operating Domains - Supplement for GNF3 Fuel. The NRC Final Safety Evaluation for this supplement (Reference 6) accepted the technical conclusions regarding applicability to GNF3 and authorized inclusion into GESTAR II via Amendment 49 (Reference 7).
NEDC-33931P Revision 2 does not change the information presented in the RAI or its associated response.
RAI-SFNB-7 The description of the analysis in NEDC-33932P Revision 1 Section 5.5.2 Normal Bias Cases provides a brief description of the analysis performed to evaluate No inserts on the rack periphery. The analysis considers perturbed scenarios referenced to a non perturbed scenario.
NEDC-33932P Revision 1 Section 5.5.3 Abnormal/Accident Bias Cases provides a brief description of the analysis performed to evaluate Abnormal positioning of fuel assembly outside the fuel storage rack. This analysis also considers perturbed scenarios referenced to a non perturbed scenario. The descriptions provided indicates the non perturbed scenario values in both the No inserts on the rack periphery and the Abnormal positioning of fuel assembly outside the fuel storage rack evaluations should be identical. However, comparison of the information listed in Table 11 to Table 12 indicates they are not identical. To evaluate compliance with 10 CFR 50.68(b)(4) explain the differences between the non perturbed scenario values in these tables.
CEG Response to RAI-SFNB-7 The No inserts on the rack periphery and Abnormal positioning of fuel assembly outside the fuel storage rack cases are not intended to be identical and as a result NEDC-33932P 6 of 8

ATTACHMENT 1 Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies (Non-Proprietary Version)
Tables 11 and 12 should not be identical. In the rack periphery study detailed in NEDC-33932P, Revision 1, Section 5.5.2, the non-perturbed case ((
                                                                                        )).
In the study examining the abnormal positioning of fuel outside of the storage rack in NEDC-33932P, Revision 1, Section 5.5.3, the non-perturbed case ((
                                                                                ))
NEDC-33931P Revision 2 does not change the information presented in this RAI or its associated response.
References
: 1. Quad Cities Nuclear Power Station Updated Final Safety Analysis Report (UFSAR),
Revision 16, dated October 2021
: 2. GESTAR II Amendment 37," dated March 2017 (ADAMS Accession Number ML17066A346 (proprietary version))
: 3. GE Licensing Topical Report, NEDE-24011-P-A-31-US, "US Supplement to General Electric Standard Application for Reactor Fuel (GESTAR II)," dated November 2020 (ADAMS Accession Numbers ML20330A195 (proprietary version) and ML20330A196 (non-proprietary version))
: 4. Letter, K. Hsueh (NRC) to J. Head (GNF-A), "Final Safety Evaluation for Amendment 37 to Global Nuclear Fuel - Americas Topical Report NEDE-24011-P-A-US General Electric Standard Application for Reactor Fuel and the US Supplement (CAC NO. MF0743)," dated March 13, 2017 (ADAMS Accession Numbers ML17066A291 and ML17069A311)
: 5. Letter, B. Moore (GNF-A) to U.S. Nuclear Regulatory Commission, "Administrative Amendment 49 to NEDE-24011-P-A-27, General Electric Standard Application for Reactor Fuel (GESTAR II)," dated October 1, 2018 (ADAMS Accession Number ML18274A195)
: 6. Letter, D. Morey (NRC) to M. Catts (GNF-A), "Final Safety Evaluation for NEDC-33173P Supplement 5 - Applicability of GE Methods to Expanded Operating Domains - Supplement for GNF3 Fuel (EPID: L-2017-TOP-0033)," dated March 21, 2019 (ADAMS Accession Numbers ML19064A229 (proprietary version) and ML19074A054 (non-proprietary version))
: 7. Letter, D. Morey (NRC) to M. Catts (GNF-A), "Final Safety Evaluation for Proposed Administrative Amendment 49 to NEDE-24011-P-A-27, General Electric Standard Application for Reactor Fuel (GESTAR II) (EPID L-2018-TOP-0039)," dated September 25, 2019 (ADAMS Accession Number ML19267A051) 7 of 8

ATTACHMENT 1 Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies (Non-Proprietary Version)
: 8. Letter 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), dated November 10, 1999 (ADAMS Accession Number ML993230184)
: 9. General Electric information Letter SIL 152, Criticality Margins for Storage of New Fuel, dated March 31, 1976, which is Attachment A to Letter from D.J. Modeen (Nuclear Energy Institute), the U.S. Nuclear Regulatory Commission, "Comments on the Criticality Accident Requirements Proposed and Direct Final Rulemaking (62 Fed. Reg. 63825 and 63911),
dated January 2, 1998 (ADAMS Accession Number ML20198C524) 8 of 8

ATTACHMENT 2 QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 Docket Nos. 50-254 and 50-265 NEDO-33932, "Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis,"
Revision 2, dated July 2022 (Non-Proprietary)

Global Nuclear Fuel NEDO-33932 Revision 2 July 2022 Non-Proprietary Information Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis Copyright 2022, 2021 Global Nuclear Fuel - Americas, LLC All Rights Reserved

NEDO-33932 Revision 2 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 Constellation Energy Generation, LLC (Constellation) and GNF, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Constellation, 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 2 Non-Proprietary Information Revision Status Revision Date                         Description of Change Number 0      May 2021       Initial release 1     October 2021    Revised marked proprietary content Updated the Section 5.5.3         description for the 2      July 2022 dropped/damaged fuel scenario.
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NEDO-33932 Revision 2 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 2 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 2 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.% .........................................................................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 2 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 2 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 2 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 2 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 2 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 2 Non-Proprietary Information Table 2 - Summary of the Critical Benchmark Experiments Experiment             Experiments      Year              Where
((
                                                                       ))
5

NEDO-33932 Revision 2 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&deg;C up to ~100&deg;C                    4-126&deg;C Moderator (in fuel region)                     H2O                            H2O Physical Form                             Solution                       Solution Temperature                        ~20&deg;C up to ~100&deg;C                    4-126&deg;C Reflector (in fuel region)                     H2O                            H2O Physical Form                             Solution                       Solution Temperature                                 20&deg;C                         4-126&deg;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. ((
                                    ))
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 2 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 2 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
((
                                                 ))
: 4. 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 2 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. ((
2.
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NEDO-33932 Revision 2 Non-Proprietary Information
                                                          ))
: 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. ((
       ))
: 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.
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NEDO-33932 Revision 2 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 2 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
((                                                                                   ))
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NEDO-33932 Revision 2 Non-Proprietary Information
((
                                                                                      ))
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 2 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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NEDO-33932 Revision 2 Non-Proprietary Information
((
                                                                            ))
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 2 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 2 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 2 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 2 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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((
                                                                                                 ))
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 2 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 2 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
((
                                                             ))
24

NEDO-33932 Revision 2 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.
25

NEDO-33932 Revision 2 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.
26

NEDO-33932 Revision 2 Non-Proprietary Information
((
                                                                                     ))
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) ((                           ))
27

NEDO-33932 Revision 2 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 2 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 2 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 2 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 2 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 2 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  kUncertaint 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 2 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 2 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 2 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)
((
36

NEDO-33932 Revision 2 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) 37

NEDO-33932 Revision 2 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) 38

NEDO-33932 Revision 2 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) 39

NEDO-33932 Revision 2 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)
                                                                                  ))
40

NEDO-33932 Revision 2 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 2 Non-Proprietary Information
((
                                                         ))
Figure 10 - Scatterplot of knorm versus 235U wt.%
((
                                                      ))
Figure 11 - Scatterplot of knorm versus 239Pu wt.%
42

NEDO-33932 Revision 2 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 2 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 2 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                                           (A-9) n 1
i 1 2
t 45

NEDO-33932 Revision 2 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 2 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 Constellation 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 2 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
((
                                                                                                ))
48

ATTACHMENT 3 QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 Docket Nos. 50-254 and 50-265 Global Nuclear Fuels - Americas, LLC 10 CFR 2.390 Affidavit for Attachment 6

Global Nuclear Fuel - Americas, LLC AFFIDAVIT I, Kent Halac, state as follows:
(1) I am the Senior Engineer, 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 the letter from P. R. Simpson (Constellation Energy Generation, LLC) to the Nuclear Regulatory Commission, RS-22-090, Response to Request for Additional Information Regarding Quad Cities New Fuel Storage Vault and Spent Fuel Storage Pool Criticality Methodologies, dated July 2022. GNF-A proprietary information in RS-22-090 is identified by a dotted underline inside double square brackets. ((This sentence is an example {3})). GNF-A proprietary information in figures and large objects is identified by double square brackets before and after the object. In each case, 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 U.S.C. &sect;552(b)(4), and the Trade Secrets Act, 18 U.S.C. &sect;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 qualifies 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 F.2d 871 (D.C. Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F.2d 1280 (D.C. Cir. 1983).
(4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a and (4)b. Some examples of categories of information that 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 a license from GNF-A constitutes a competitive economic advantage over other companies;
: b. Information that, if used by a competitor, would reduce its expenditure of resources or improve its competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
: c. Information that reveals aspects of past, present, or future GNF-A customer-funded development plans and programs, resulting in potential products to GNF-A; RS-22-090                                                                      Affidavit Page 1 of 3

Global Nuclear Fuel - Americas, LLC
: d. Information that discloses trade secret or potentially patentable subject matter for which it may be desirable to obtain patent protection.
(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, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions for proprietary or confidentiality agreements or both that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in the following paragraphs (6) and (7).
(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be 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 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 and/or confidentiality agreements.
(8) The information identified in paragraph (2) is classified as proprietary because it contains the detailed GNF-A methodology for fuel analyses for the GNF-A Boiling Water Reactor (BWR). These methods, techniques, and data along with their application to the design, modification, and analyses associated with the fuel analyses were achieved at a significant cost to GNF-A.
The development of the evaluation processes along with the interpretation and application of the analytical results is derived from the extensive experience databases that constitute 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 RS-22-090                                                                      Affidavit Page 2 of 3

Global Nuclear Fuel - Americas, LLC 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.
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 there 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 7th day of July 2022.
Kent Halac Senior Engineer, Regulatory Affairs Global Nuclear Fuels - Americas, LLC 3901 Castle Hayne Road Wilmington, NC 28401 Kent.Halac@ge.com RS-22-090                                                                      Affidavit Page 3 of 3

ATTACHMENT 4 QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 Docket Nos. 50-254 and 50-265 Global Nuclear Fuels - Americas, LLC 10 CFR 2.390 Affidavit for Attachment 7

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 2, July 2022. GNF-A proprietary information within the text and tables is identified by a dotted underline placed within double square brackets.
((This sentence is an example.{3})) Figures and large objects containing GNF-A proprietary information are identified with double square brackets before and after the object. In all cases, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.
(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GNF-A relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F2d 871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F2d 1280 (DC Cir. 1983).
(4) Some examples of categories of information which fit into the definition of proprietary information are:
: a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GNF-A's competitors without license from GNF-A constitutes a competitive economic advantage over other companies;
: b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
: c. Information which reveals aspects of past, present, or future GNF-A customer-funded development plans and programs, resulting in potential products to GNF-A;
: d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
NEDC-33932P Revision 2                                                  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 2                                                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 1st day of July 2022.
Brian R. Moore General Manager, Core & Fuel Engineering Global Nuclear Fuel - Americas, LLC 3901 Castle Hayne Road Wilmington, NC 28401 Brian.Moore@ge.com NEDC-33932P Revision 2                                                  Affidavit Page 3 of 3

ATTACHMENT 5 QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 Docket Nos. 50-254 and 50-265 Curtiss-Wright Corporation 10 CFR 2.390 Affidavit for Attachment 7

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 2, July 2022.
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.
Harris, Matt 2022.07.06 10:33:14 -04'00' Matthew C. Harris Segment Manager, NETCO , business segment of Scientech Curtiss-Wright Corporation Date: __7/6/22__________
Page 2 of 2 Curtiss Wright Nuclear Division 44 Shelter Rock Rd
* Danbury, CT 06810
* Phone: 203.448.3439
* Fax: 203.437.6279}}