RS-15-326, Request for License Amendment Regarding Spent Fuel Storage Pool Criticality Methodology for Fuel Channel Bow/Bulge

Request for License Amendment Regarding Spent Fuel Storage Pool Criticality Methodology for Fuel Channel Bow/Bulge

ML15348A396
Person / Time
Site:	Quad Cities
Issue date:	12/14/2015
From:	Simpson P R Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15348A395	List: RS-15-326, Request for License Amendment Regarding Spent Fuel Storage Pool Criticality Methodology for Fuel Channel Bow/Bulge
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RS-15-326
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4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office Attachments 2 and 3 contain Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachments 2 and 3, this document is decontrolled. Proprietary Information - Withhold From Public Disclosure Under 10 CFR 2.390 RS-15-326 10 CFR 50.90 December 14, 2015

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

Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

Request for License Amendment Regarding Spent Fuel Storage Pool Criticality Methodology for Fuel Channel Bow/Bulge

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, respectively. The proposed change would allow use of a new Criticality Safety Analysis (CSA) fuel channel bow/bulge methodology for performing the criticality safety evaluation for the new ATRIUM 10XM fuel design in the spent fuel pool. A description and evaluation of the proposed change is attached. This license amendment request was discussed with the NRC in a pre-application meeting on May 11, 2015. During the meeting, the NRC questioned the treatment of tolerances and uncertainties in the analysis. To address the concerns, the NRC requested EGC to include more detail concerning the treatment of tolerances and uncertainties. Accordingly, the requested information is attached. This request is subdivided as follows.

• Attachment 1 provides a description and evaluation of the proposed change.

• Attachment 2 provides AREVA, Inc. Report FS1-0024092, Revision 1.0, "Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Proprietary Version."

• Attachment 3 provides a comparison of the approach used in the NRC-approved SVEA-96 Optima2 CSA to the approach used in the proposed ATRIUM 10XM CSA, as well as a comparison of the SVEA-96 Optima2 and ATRIUM 10XM biases and uncertainties. This information was requested by the NRC in a pre-application meeting that was held on May 11, 2015.

December 14, 2015 U.S. Nuclear Regulatory Commission

Page 3 3. Table 1: Criticality Safety Analysis (CSA) Approach Comparison, and Table 2: Comparison of the ATRIUM 10XM and SVEA-96 Optima2 Biases and Uncertainties (Proprietary Version) 4. AREVA, Inc. Affidavit

5. AREVA, Inc. Report FS1-0024106, Revision 1.0, "Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version" 6. Holtec International Affidavit

7. Table 1: Criticality Safety Analysis (CSA) Approach Comparison, and Table 2: Comparison of the ATRIUM 10XM and SVEA-96 Optima2 Biases and Uncertainties (Nonproprietary Version)

cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector - Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety ATTACHMENT 1 Evaluation of Proposed Change Page 1 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable

Regulatory Requirements/Criteria 4.2 No Significant Hazards Consideration

6.0 REFERENCES

ATTACHMENT 1 Evaluation of Proposed Change Page 2 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power

Station (QCNPS), Units 1 and 2, respectively. The proposed change would allow use of a new Criticality Safety Analysis (CSA) fuel channel bow/bulge methodology for performing the criticality safety evaluation for the new ATRIUM 10XM (A10XM) fuel design in the spent fuel

pool (SFP). 2.0 DETAILED DESCRIPTION EGC is planning to transition from Westinghouse SVEA-96 Optima2 fuel to the new AREVA A10XM fuel design at QCNPS in spring 2017. Based on the EGC plans to transition to a new fuel design, Holtec International revised the SFP CSA to account for the new A10XM fuel.

While this revised SFP CSA supports the planned transition to A10XM fuel, this new analysis is not required to support the NRC review and approval of the separate fuel transition amendment request submitted on February 6, 2015 (i.e., Reference 1). In Reference 2, the NRC issued Amendment No. 253 to Renewed Facility Operating License No. DPR-29 and Amendment No. 248 to Renewed Facility Operating License No. DPR-30 for QCNPS Units 1 and 2, respectively. These amendments established the SFP CSA methodologies for QCNPS. The revised SFP CSA for A10XM fuel has been performed in accordance with these NRC reviewed and approved methodologies, with the exception of the treatment of fuel channel bow/bulge. The NRC Safety Evaluation associated with Amendment Nos. 253 and 248 stated: While the methodology evaluated by NRC staff in the NCS analyses submitted for review was found to be acceptable, the methodology did omit details on the appropriate approach to use in evaluation of fuel channel bowing/bulging or reconstituted fuel. The licensee provided a satisfactory explanation for not including these conditions in the current analysis, but declined to provide a detailed explanation of how these conditions might be modeled (including potential conservatisms/non-conservatisms, uncertainties, and biases). Therefore, the findings in this safety evaluation do not extend to such conditions.

Consequently, the fuel channel bow/bulge methodology used in the revised SFP CSA for A10XM fuel is submitted at this time for NRC review and approval.

3.0 TECHNICAL EVALUATION

QCNPS Updated Final Safety Analysis Report (UFSAR) Section 9.1.2 documents the QCNPS, Units 1 and 2 SFP safety design bases as summarized below. The similarity of the SFP storage rack designs for the two units permit a single set of supporting analyses to apply to both units. The spent fuel assembly racks, with NETCO-SNAP-IN rack inserts, are designed to ensure subcriticality in the storage pool. A maximum k eff of 0.95 is maintained with the racks fully ATTACHMENT 1 Evaluation of Proposed Change Page 3 loaded with fuel of the highest anticipated r eactivity and flooded with unborated water at a temperature corresponding to the highest reactivity. The spent fuel storage pools have been designed to withstand the anticipated earthquake loadings as a Class I structure. The high-density racks are engineered to achieve the dual objective of maximum protection against structural loading (such as ground motion) and the maximization of available storage locations. There are various legacy fuel assembly designs, including the current SVEA-96 Optima2 design that are qualified for storage in the QCNPS Units 1 and 2 SFPs as documented in QCNPS UFSAR Section 9.1.2.3. EGC is planning to transition from Westinghouse SVEA-96 Optima2 fuel to the AREVA A10XM fuel design at QCNPS. To support future operations, the A10XM fuel assembly is designed to be compatible with the QCNPS reactor core and co-resident legacy fuel. The A10XM fuel assembly is constructed of similar materials within a spatial envelope that is similar to the currently licensed SVEA-96 Optima2 legacy fuel type. The A10XM design must be qualified for storage in the QCNPS Units 1 and 2 SFPs. A CSA for the QCNPS Units 1 and 2 SFPs has been performed to support the planned transition to A10XM fuel. The A10XM CSA uses the CSA methodology approved in Reference 2 with the exception of the fuel channel bow/bulge treatment. Specifically, the A10XM CSA was performed using the computer codes CASMO-4 and MCNP5, a peak reactivity lattice, a minimum Boron-10 areal density in the QCNPS SFP rack inserts of

0.0116 g/cm 2, and incorporating fuel assembly and storage rack manufacturing tolerances consistent with the CSA approved in Reference 2. Attachment 3, Table 1, provides a comparison of the A10XM CSA and the CSA approaches approved in Reference 2. This comparison was requested by the NRC in a pre-application meeting that was held on May 11, 2015. Fuel channel bulging and bowing is a depletion related geometry change that changes the proximity of the channel to the fuel rods. The volume of moderator inside the channel is impacted by this change in distance from the channel to fuel rods. Thus, the fuel to moderator ratio may change during depletion, as a result of fuel channel bulging and bowing. A fuel channel bow/bulge bias and uncertainty is not included in the total biases and uncertainties for the A10XM fuel (see Attachment 3, Table 2) because the peak reactivity of the A10XM lattices is at about 10-15 GWD/MTU. As shown in Attachment 2, Figures 1 and 3, fuel geometry changes are not expected to occur for the A10XM fuel design at such low exposures.

Beyond this exposure range, the fuel channel bow/bulge does increase. However, the reactivity increase from this geometry change is offset by the decrease in reactivity of the fuel with exposure. The results of the A10XM SFP CSA demonstrate that the total biases and uncertainties for the A10XM fuel are less than the total biases and uncertainties for the SVEA-96 Optima2 fuel (see ATTACHMENT 1 Evaluation of Proposed Change Page 4 Attachment 3, Table 2). Therefore, the k eff of an A10XM assembly is bounded by the k eff of an equivalent reactivity SVEA-96 Optima2 assembly. The CSA for the storage of A10XM assemblies in the QCNPS spent fuel storage racks with NETCO-SNAP-IN rack inserts has been performed. The results for the normal condition show that k eff is < 0.95, with a 95 percent probability at a 95 percent confidence level, with the storage racks fully loaded with A10XM fuel at the Technical Specifications 4.3.1.1.c maximum in-rack k-infinity limit. The results for the bounding accident condition (i.e., missing insert with centric fuel positioning) also show that k eff is < 0.95, with a 95 percent probability at a 95 percent confidence level, with the storage racks fully loaded with A10XM fuel at the Technical Specifications 4.3.1.1.c maximum in-rack k-infinity limit.

4.0 REGULATORY EVALUATION

4.1 Applicable

Regulatory Requirements/Criteria 10 CFR 50.68, "Criticality accident requirements," paragraph (b)(4) states that the k eff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. A SFP CSA has been performed to demonstrate that this

requirement is met. Paragraph (b)(7) of 10 CFR 50.68 states that the maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to 5.0 percent by weight. QCNPS new fuel is below 5.0 percent by weight U-235 enrichment. The following General Design Criterion (GDC) is applicable to this amendment request. It should be noted that, although QCNPS is not formally committed to the GDC due to the vintage of the station, an evaluation was performed addressing the QCNPS conformance with the GDC. This evaluation is documented in the UFSAR Section 3.1, "Conformance with NRC General Design Criteria." This evaluation concluded that QCNPS fully satisfies the intent of the (then draft) GDC. GDC 5, "Sharing of structures, systems, and components," specifies that 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 of the remaining units. The spent fuel storage pool has been designed to withstand the anticipated earthquake loadings as a Class I structure.

Each unit has its own SFP measuring 33 x 41 feet. The fuel storage pools of Units 1 and 2 are connected by a double-gated transfer canal. The fuel pool is a reinforced-concrete structure, lined with seam-welded, stainless steel plate, welded to reinforcing members embedded in concrete. The 3/16-inch stainless steel liner will prevent leakage in the unlikely event the concrete develops cracks. To avoid unintentional draining of the pool, there are no penetrations that would permit the pool to be drained below a safe storage level. The passage between the fuel storage pool and the reactor cavity is located above the reactor vessel, is constructed with two, double-sealed gates and has a monitored drain between the gates. This arrangement permits detection of leaks from ATTACHMENT 1 Evaluation of Proposed Change Page 5 the passage and repair of a leaking gate. The depth of water in the fuel storage pool is approximately 37 feet 9 inches and the depth of the water in the transfer canal during refueling is 22 feet 9 inches. The proposed change only modifies the methodology for treating fuel channel bow/bulge in the SFP CSA; therefore, compliance with GDC 5 is not affected by the proposed change. GDC 62, "Prevention of criticality in fuel storage and handling," states that criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations. The evaluation of QCNPS's conformance with GDC 62 is discussed in Section 9.1.2, "Spent Fuel Storage," of the QCNPS UFSAR. A SFP CSA has been performed to demonstrate that k eff will remain less than or equal to 0.95 while accounting for potential A10XM fuel channel bow/bulge. Therefore, compliance with GDC 62 is not affected by the proposed change. 4.2 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, respectively. The proposed change would allow use of a new Criticality Safety Analysis (CSA) fuel channel bow/bulge methodology for performing the criticality safety evaluation for the new ATRIUM 10XM fuel design in the spent fuel pool (SFP). According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in the probability or consequences of any accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

EGC has evaluated the proposed change, using the criteria in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

ATTACHMENT 1 Evaluation of Proposed Change Page 6 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No The proposed change involves a revised CSA for the QCNPS Units 1 and 2 SFPs using a new fuel channel bow/bulge methodology. The proposed change does not alter or modify the fuel, fuel handling processes, spent fuel storage racks, number of fuel assemblies that may be stored in the SFP, decay heat generation rate, or the SFP cooling and cleanup system. The proposed change was evaluated for impact on the following previously evaluated events and accidents:

• A fuel handling accident (FHA),

• A fuel mispositioning event,

• A seismic event, and

• A loss of SFP cooling event. The probability of a FHA is not increased because implementation of the proposed change will employ the same equipment and processes to handle fuel assemblies that are currently used. The FHA radiological consequences are not increased because the fuel channel bow/bulge methodology used in the CSA does not impact the radiological source term of a single fuel assembly. Therefore, the proposed change does not significantly increase the probability or consequences of an FHA. Operation in accordance with the proposed change will not significantly increase the probability of a fuel mispositioning event because fuel movement will continue to be controlled by approved fuel handling procedures. These procedures continue to require identification of the initial and target locations for each fuel assembly that is moved. The consequences of a fuel mispositioning event are not changed because the reactivity analysis demonstrates that the new subcriticality criteria and requirements will be met for the worst-case fuel

mispositioning event. Operation in accordance with the proposed change will not change the probability of a seismic event. The consequences of a seismic event are not increased because the forcing functions for seismic excitation are not increased and because the mass of storage racks has not changed. Operation in accordance with the proposed change will not change the probability of a loss of SFP cooling event because the systems and events that could affect SFP cooling are unchanged. The consequences are not significantly increased because there are no changes in the SFP heat load or SFP cooling systems, structures or components due to the proposed change in fuel channel bow/bulge methodology used in the CSA.

ATTACHMENT 1 Evaluation of Proposed Change Page 7 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No Onsite storage of spent fuel assemblies in the QCNPS, Units 1 and 2, SFPs is a normal activity for which QCNPS has been designed and licensed. As part of assuring that this normal activity can be performed without endangering the public health and safety, the ability to safely accommodate different possible accidents in the spent fuel pool have been previously analyzed. These analyses address accidents such as radiological releases due to dropping a fuel assembly; and potential inadvertent criticality due to misloading a fuel assembly. The proposed change does not alter the method of fuel movement or spent fuel storage and does not create the potential for a new accident. The proposed use of a new fuel channel bow/bulge methodology for performing the QCNPS revised SFP CSA does not change or modify the fuel, fuel handling processes, spent fuel racks, number of fuel assemblies that may be stored in the pool, decay heat generation rate, or the SFP cooling and cleanup system. The limiting fuel assembly mispositioning event does not represent a new or different type of accident. The mispositioning of a fuel assembly within the fuel storage racks has always been possible. The proposed change involves a revised CSA for the QCNPS, Units 1 and 2, SFPs using a new fuel channel bow/bulge methodology. The associated analysis results show that the storage racks remain sub-critical, with substantial margin, following a worst-case fuel

misloading event. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No The proposed change involves a revised CSA for the QCNPS, Units 1 and 2, SFPs using a new fuel channel bow/bulge methodology. This change was evaluated for its effect on margins of safety related to criticality and spent fuel heat removal capability. QCNPS Technical Specifications Section 4.3, "Fuel Storage," Specification 4.3.1.1.a requires the spent fuel storage racks to maintain the effective neutron multiplication factor, keff, less than or equal to 0.95 when fully flooded with unborated water, which includes an allowance for uncertainties. Therefore, for SFP criticality considerations, the required safety margin is five percent.

ATTACHMENT 1 Evaluation of Proposed Change Page 8 The proposed change ensures, as verified by the associated criticality analysis, that k eff continues to be less than or equal to 0.95, thus preserving the required safety margin of five percent. The proposed use of a new fuel channel bow/bulge methodology for performing the QCNPS SFP CSA does not affect spent fuel heat generation or the spent fuel cooling systems. In addition, the radiological consequences of a dropped fuel assembly remain unchanged as the anticipated fuel damage due to a fuel handling accident is unaffected by the use of a new fuel channel bow/bulge methodology to perform the CSA. The proposed change also does not increase the capacity of the Unit 1 and Unit 2 spent fuel pools beyond the current capacity of no more than 3657 and 3897 fuel assemblies, respectively. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above evaluation, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, paragraph (c), and accordingly, a finding of no significant hazards consideration is justified.

4.3 Conclusions

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

5.0 ENVIRONMENTAL CONSIDERATION

EGC has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation." However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review,"

paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

ATTACHMENT 1 Evaluation of Proposed Change Page 9

6.0 REFERENCES

1. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "Request for License Amendment Regarding Transition to AREVA Fuel," dated February 6, 2015 2. Letter from B. Mozafari (U.S. NRC) to M. J. Pacilio (Exelon Generation Company, LLC), "Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments Regarding NETCO Inserts (TAC Nos. MF2489 and MF2490)(RS-13-148)," dated December 31, 2014 ATTACHMENT 4 AREVA, Inc. Affidavit

ATTACHMENT 5 AREVA, Inc. Report FS1-0024106, Revision 1.0, "Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version"

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 2/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page REVISIONS REVISION DATE EXPLANATORY NOTES 1.0 See 1 st page release date New document. This is the Nonproprietary version of FS1-0024092.

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 3/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page TABLE OF CONTENTS

1. PURPOSE .......................................................................................................................

................. 4

2. METHODOLOGY/APPROACH ........................................................................................................ 5

3. ASSUMPTIONS ...................................................................................................................

............ 6

3.1. ASSUMPTIONS

...................................................................................................................

.................... 6

3.2. ASSUMPTIONS

REQUIRING VERIFICATION ....................................................................................... 6

4. DETERMINATION OF VALUES ......................................................................................................

7 4.1. EQUIVALENT FAST FLUENCE .......................................................................................................

....... 7 4.2. EXPECTED FUEL ROD GROWTH ................

......................................................................................

... 7 4.3. CLADDING CREEP ................................................................................................................

................. 7

4.4. CHANNEL

BOW AND BULGE .........................................................................................................

....... 7 5. REFERENCES ....................................................................................................................

........... 13 LIST OF TABLES Table 1 Assembly Fast Fluence by Exposure ....................................................................................

.........

7 LIST OF FIGURES Figure 1 Mean Bulge by Exposure for Different Channel Geometries .........................................................

9 Figure 2 Bulge Standard Deviations by Exposure for Different Channel Geometries ...............................

10 Figure 3 Mean Bow by Exposure for Different Channel Geometries .........................................................

11 Figure 4 Bow Standard Deviations by Exposure for Different Channel Geometries ................................. 12

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 4/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page

1. PURPOSE As a part of the licensing of ATRIUM 10XM fuel assemblies for the Dresden and Quad Cities units, the customer, EXELON, has had spent fuel rack criticality analyses performed by another vendor. The submittal of these analyses to the NRC has resulted in two Requests for Additional Information (RAIs) related to AREVA fuel assembly and fuel channel performance. These RAIs read as follows: 11. Sections 2.3.11.1.1 and 2.3.11.1.2 of the HI-2146153 analysis explain that fuel rod growth, cladding creep, and crud buildup do not need to be evaluated because these factors are not expected to be significant at the peak reactivity burnup of the design basis lattice. Changes to the fuel rod geometry as a result of irradiation may result in a positive reactivity impact. Provide information regarding the expected fuel rod growth, cladding creep, and crud buildup at this burnup, and explain why the reactivity impact would not be significant.

13. Section 2.3.11.2 of the HI-2146153 analysis does not describe clearly how the geometry is changed to evaluate fuel channel bulging and bowing. The text refers to the channel outer exposed width tolerance, but it is not clear if the outer exposed width is changed by varying the channel inner width or the channel wall thickness.

Describe how the MCNP model was altered for Case 2.3.11.2.1. Also, discuss how the channel bowing tolerance was determined and how it bounds any expected ATRIUM 10XM channel bulging/bowing.

The customer has requested that such information be provided for fuel assembly exposures in the ranges of 10-20 GWd/MTU and 40-50 GWd/MTU. Specific AREVA information requested to support responses to these RAIs are: 1) expected fuel rod growth; 2) cladding creep; 3) crud buildup; and 4) channel bowing and bulge. For fuel rod crud buildup, such data are core-dependent and must be supplied by the customer.

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 5/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page

2. METHODOLOGY/APPROACH This Engineering Information Record provides values responsive to the requests described in §1 based on design bases related to fuel channel design or design basis data for the RODEX4 methodology, which is used to analyze the thermal-mechanical performance of ATRIUM 10XM fuel rods.

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 6/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page

3. ASSUMPTIONS 3.1. ASSUMPTIONS There are no assumptions made in this EIR. 3.2. ASSUMPTIONS REQ UIRING VERIFICATION There are no assumptions requiring verification as all values are obtained from quality records.

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 7/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page

4. DETERMINATION OF VALUES 4.1. EQUIVALENT FAST FLUENCE Since the growth of the Zircaloy comprising the fuel rods is well correlated with fast neutron fluence, mechanistic or correlative models of fuel rod growth and cladding creep are provided as a function of this fluence. Since the customer has requested the information described above as a function of assembly-averaged exposure, it is necessary to provide fast fluence ranges that correspond to those exposure ranges. Approximate values are shown in Table 1 [2, Fig. 2].

Table 1 Assembly Fast Fluence by Exposure

4.2. EXPECTED

FUEL ROD GROWTH For expected fuel rod growth, which includes both growth and PCMI-induced creep, appropriate values in are found in the response to RAI #3 in the first round of RAIs to the RODEX4 topical report [1, BAW-10247Q1(P), p. 4]. Upper limits for both of the exposure ranges, based on the BWR data compilation are

[ ]

for the 10 GWd/MTU to 20 GWd/MTU range, and

[ ] for the 40 GWd/MTU to 50 GWd/MTU range. 4.3. CLADDING CREEP Cladding creep, manifesting itself in diametral reduction is shown the RODEX4 topical report [1, Figure 4.16]. For the first exposure interval of 10 GWd/MTU to 20 GWd/MTU, the maximum reduction is less than

[ ], while for the second exposure interval of 40 GWd/MTU to 50 GWd/MTU, the maximum reduction is bounded by

[ ]. 4.4. CHANNEL BOW AND BULGE The RODEX4 methodology does not consider channel bulge as bulge occurs predominantly at axial locations where there is little or no in-channel voiding. In addition, the RODEX4 methodology places no limits on channel bowing, but explicitly computes bowing based on fast fluence gradients across the channel and accordingly penalizes linear power margins. As a part of the channel mechanical design methodology, channel bow and bulge statistics are provided as inputs to a Monte Carlo methodology for

assessing interference with control blades [3, §7.0]. Plots of these statistics are provided as function of exposure and lattice geometry in Figure 1 through Figure 4 [3, Figures 7.7 through 7.10]. The values denoted as D-lattice 2.54 mm correspond to 100 mil D-lattice channels, while those denoted as D-lattice 2.10 mm correspond to 80 mil D-lattice channels.

[

]

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 8/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page

[ ] reasonably bounds the observed population of channels for each exposure range.

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 9/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page Figure 1 Mean Bulge by Exposure for Different Channel Geometries

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 10/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page Figure 2 Bulge Standard Deviations by Exposure for Different Channel Geometries

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 11/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page Figure 3 Mean Bow by Exposure for Different Channel Geometries

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 12/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page Figure 4 Bow Standard Deviations by Exposure for Different Channel Geometries

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 13/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page

5. REFERENCES 1 BAW-10247PA, Revision 0. Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors. April 2008.

2 [ ] 3 EMF-93-177(P)(A), Revision 1.

Mechanical Design for BWR Fuel Channels.

August 2005.

ATTACHMENT 6 Holtec International Affidavit

ATTACHMENT 7 Table 1: Criticality Safety Analysis (CSA) Approach Comparison, and Table 2: Comparison of the ATRIUM 10XM and SVEA-96 Optima2 Biases and Uncertainties (Nonproprietary Version)

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office Attachments 2 and 3 contain Proprietary Information. Withhold from public disclosure under 10 CFR 2.390. When separated from Attachments 2 and 3, this document is decontrolled. Proprietary Information - Withhold From Public Disclosure Under 10 CFR 2.390 RS-15-326 10 CFR 50.90 December 14, 2015

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

Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

Request for License Amendment Regarding Spent Fuel Storage Pool Criticality Methodology for Fuel Channel Bow/Bulge

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, respectively. The proposed change would allow use of a new Criticality Safety Analysis (CSA) fuel channel bow/bulge methodology for performing the criticality safety evaluation for the new ATRIUM 10XM fuel design in the spent fuel pool. A description and evaluation of the proposed change is attached. This license amendment request was discussed with the NRC in a pre-application meeting on May 11, 2015. During the meeting, the NRC questioned the treatment of tolerances and uncertainties in the analysis. To address the concerns, the NRC requested EGC to include more detail concerning the treatment of tolerances and uncertainties. Accordingly, the requested information is attached. This request is subdivided as follows.

• Attachment 1 provides a description and evaluation of the proposed change.

• Attachment 2 provides AREVA, Inc. Report FS1-0024092, Revision 1.0, "Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Proprietary Version."

• Attachment 3 provides a comparison of the approach used in the NRC-approved SVEA-96 Optima2 CSA to the approach used in the proposed ATRIUM 10XM CSA, as well as a comparison of the SVEA-96 Optima2 and ATRIUM 10XM biases and uncertainties. This information was requested by the NRC in a pre-application meeting that was held on May 11, 2015.

December 14, 2015 U.S. Nuclear Regulatory Commission

Page 3 3. Table 1: Criticality Safety Analysis (CSA) Approach Comparison, and Table 2: Comparison of the ATRIUM 10XM and SVEA-96 Optima2 Biases and Uncertainties (Proprietary Version) 4. AREVA, Inc. Affidavit

5. AREVA, Inc. Report FS1-0024106, Revision 1.0, "Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version" 6. Holtec International Affidavit

7. Table 1: Criticality Safety Analysis (CSA) Approach Comparison, and Table 2: Comparison of the ATRIUM 10XM and SVEA-96 Optima2 Biases and Uncertainties (Nonproprietary Version)

cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector - Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety ATTACHMENT 1 Evaluation of Proposed Change Page 1 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable

Regulatory Requirements/Criteria 4.2 No Significant Hazards Consideration

6.0 REFERENCES

ATTACHMENT 1 Evaluation of Proposed Change Page 2 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power

Station (QCNPS), Units 1 and 2, respectively. The proposed change would allow use of a new Criticality Safety Analysis (CSA) fuel channel bow/bulge methodology for performing the criticality safety evaluation for the new ATRIUM 10XM (A10XM) fuel design in the spent fuel

pool (SFP). 2.0 DETAILED DESCRIPTION EGC is planning to transition from Westinghouse SVEA-96 Optima2 fuel to the new AREVA A10XM fuel design at QCNPS in spring 2017. Based on the EGC plans to transition to a new fuel design, Holtec International revised the SFP CSA to account for the new A10XM fuel.

While this revised SFP CSA supports the planned transition to A10XM fuel, this new analysis is not required to support the NRC review and approval of the separate fuel transition amendment request submitted on February 6, 2015 (i.e., Reference 1). In Reference 2, the NRC issued Amendment No. 253 to Renewed Facility Operating License No. DPR-29 and Amendment No. 248 to Renewed Facility Operating License No. DPR-30 for QCNPS Units 1 and 2, respectively. These amendments established the SFP CSA methodologies for QCNPS. The revised SFP CSA for A10XM fuel has been performed in accordance with these NRC reviewed and approved methodologies, with the exception of the treatment of fuel channel bow/bulge. The NRC Safety Evaluation associated with Amendment Nos. 253 and 248 stated: While the methodology evaluated by NRC staff in the NCS analyses submitted for review was found to be acceptable, the methodology did omit details on the appropriate approach to use in evaluation of fuel channel bowing/bulging or reconstituted fuel. The licensee provided a satisfactory explanation for not including these conditions in the current analysis, but declined to provide a detailed explanation of how these conditions might be modeled (including potential conservatisms/non-conservatisms, uncertainties, and biases). Therefore, the findings in this safety evaluation do not extend to such conditions.

Consequently, the fuel channel bow/bulge methodology used in the revised SFP CSA for A10XM fuel is submitted at this time for NRC review and approval.

3.0 TECHNICAL EVALUATION

QCNPS Updated Final Safety Analysis Report (UFSAR) Section 9.1.2 documents the QCNPS, Units 1 and 2 SFP safety design bases as summarized below. The similarity of the SFP storage rack designs for the two units permit a single set of supporting analyses to apply to both units. The spent fuel assembly racks, with NETCO-SNAP-IN rack inserts, are designed to ensure subcriticality in the storage pool. A maximum k eff of 0.95 is maintained with the racks fully ATTACHMENT 1 Evaluation of Proposed Change Page 3 loaded with fuel of the highest anticipated r eactivity and flooded with unborated water at a temperature corresponding to the highest reactivity. The spent fuel storage pools have been designed to withstand the anticipated earthquake loadings as a Class I structure. The high-density racks are engineered to achieve the dual objective of maximum protection against structural loading (such as ground motion) and the maximization of available storage locations. There are various legacy fuel assembly designs, including the current SVEA-96 Optima2 design that are qualified for storage in the QCNPS Units 1 and 2 SFPs as documented in QCNPS UFSAR Section 9.1.2.3. EGC is planning to transition from Westinghouse SVEA-96 Optima2 fuel to the AREVA A10XM fuel design at QCNPS. To support future operations, the A10XM fuel assembly is designed to be compatible with the QCNPS reactor core and co-resident legacy fuel. The A10XM fuel assembly is constructed of similar materials within a spatial envelope that is similar to the currently licensed SVEA-96 Optima2 legacy fuel type. The A10XM design must be qualified for storage in the QCNPS Units 1 and 2 SFPs. A CSA for the QCNPS Units 1 and 2 SFPs has been performed to support the planned transition to A10XM fuel. The A10XM CSA uses the CSA methodology approved in Reference 2 with the exception of the fuel channel bow/bulge treatment. Specifically, the A10XM CSA was performed using the computer codes CASMO-4 and MCNP5, a peak reactivity lattice, a minimum Boron-10 areal density in the QCNPS SFP rack inserts of

0.0116 g/cm 2, and incorporating fuel assembly and storage rack manufacturing tolerances consistent with the CSA approved in Reference 2. Attachment 3, Table 1, provides a comparison of the A10XM CSA and the CSA approaches approved in Reference 2. This comparison was requested by the NRC in a pre-application meeting that was held on May 11, 2015. Fuel channel bulging and bowing is a depletion related geometry change that changes the proximity of the channel to the fuel rods. The volume of moderator inside the channel is impacted by this change in distance from the channel to fuel rods. Thus, the fuel to moderator ratio may change during depletion, as a result of fuel channel bulging and bowing. A fuel channel bow/bulge bias and uncertainty is not included in the total biases and uncertainties for the A10XM fuel (see Attachment 3, Table 2) because the peak reactivity of the A10XM lattices is at about 10-15 GWD/MTU. As shown in Attachment 2, Figures 1 and 3, fuel geometry changes are not expected to occur for the A10XM fuel design at such low exposures.

Beyond this exposure range, the fuel channel bow/bulge does increase. However, the reactivity increase from this geometry change is offset by the decrease in reactivity of the fuel with exposure. The results of the A10XM SFP CSA demonstrate that the total biases and uncertainties for the A10XM fuel are less than the total biases and uncertainties for the SVEA-96 Optima2 fuel (see ATTACHMENT 1 Evaluation of Proposed Change Page 4 Attachment 3, Table 2). Therefore, the k eff of an A10XM assembly is bounded by the k eff of an equivalent reactivity SVEA-96 Optima2 assembly. The CSA for the storage of A10XM assemblies in the QCNPS spent fuel storage racks with NETCO-SNAP-IN rack inserts has been performed. The results for the normal condition show that k eff is < 0.95, with a 95 percent probability at a 95 percent confidence level, with the storage racks fully loaded with A10XM fuel at the Technical Specifications 4.3.1.1.c maximum in-rack k-infinity limit. The results for the bounding accident condition (i.e., missing insert with centric fuel positioning) also show that k eff is < 0.95, with a 95 percent probability at a 95 percent confidence level, with the storage racks fully loaded with A10XM fuel at the Technical Specifications 4.3.1.1.c maximum in-rack k-infinity limit.

4.0 REGULATORY EVALUATION

4.1 Applicable

Regulatory Requirements/Criteria 10 CFR 50.68, "Criticality accident requirements," paragraph (b)(4) states that the k eff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. A SFP CSA has been performed to demonstrate that this

requirement is met. Paragraph (b)(7) of 10 CFR 50.68 states that the maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to 5.0 percent by weight. QCNPS new fuel is below 5.0 percent by weight U-235 enrichment. The following General Design Criterion (GDC) is applicable to this amendment request. It should be noted that, although QCNPS is not formally committed to the GDC due to the vintage of the station, an evaluation was performed addressing the QCNPS conformance with the GDC. This evaluation is documented in the UFSAR Section 3.1, "Conformance with NRC General Design Criteria." This evaluation concluded that QCNPS fully satisfies the intent of the (then draft) GDC. GDC 5, "Sharing of structures, systems, and components," specifies that 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 of the remaining units. The spent fuel storage pool has been designed to withstand the anticipated earthquake loadings as a Class I structure.

Each unit has its own SFP measuring 33 x 41 feet. The fuel storage pools of Units 1 and 2 are connected by a double-gated transfer canal. The fuel pool is a reinforced-concrete structure, lined with seam-welded, stainless steel plate, welded to reinforcing members embedded in concrete. The 3/16-inch stainless steel liner will prevent leakage in the unlikely event the concrete develops cracks. To avoid unintentional draining of the pool, there are no penetrations that would permit the pool to be drained below a safe storage level. The passage between the fuel storage pool and the reactor cavity is located above the reactor vessel, is constructed with two, double-sealed gates and has a monitored drain between the gates. This arrangement permits detection of leaks from ATTACHMENT 1 Evaluation of Proposed Change Page 5 the passage and repair of a leaking gate. The depth of water in the fuel storage pool is approximately 37 feet 9 inches and the depth of the water in the transfer canal during refueling is 22 feet 9 inches. The proposed change only modifies the methodology for treating fuel channel bow/bulge in the SFP CSA; therefore, compliance with GDC 5 is not affected by the proposed change. GDC 62, "Prevention of criticality in fuel storage and handling," states that criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations. The evaluation of QCNPS's conformance with GDC 62 is discussed in Section 9.1.2, "Spent Fuel Storage," of the QCNPS UFSAR. A SFP CSA has been performed to demonstrate that k eff will remain less than or equal to 0.95 while accounting for potential A10XM fuel channel bow/bulge. Therefore, compliance with GDC 62 is not affected by the proposed change. 4.2 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, respectively. The proposed change would allow use of a new Criticality Safety Analysis (CSA) fuel channel bow/bulge methodology for performing the criticality safety evaluation for the new ATRIUM 10XM fuel design in the spent fuel pool (SFP). According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in the probability or consequences of any accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

EGC has evaluated the proposed change, using the criteria in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

ATTACHMENT 1 Evaluation of Proposed Change Page 6 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No The proposed change involves a revised CSA for the QCNPS Units 1 and 2 SFPs using a new fuel channel bow/bulge methodology. The proposed change does not alter or modify the fuel, fuel handling processes, spent fuel storage racks, number of fuel assemblies that may be stored in the SFP, decay heat generation rate, or the SFP cooling and cleanup system. The proposed change was evaluated for impact on the following previously evaluated events and accidents:

• A fuel handling accident (FHA),

• A fuel mispositioning event,

• A seismic event, and

• A loss of SFP cooling event. The probability of a FHA is not increased because implementation of the proposed change will employ the same equipment and processes to handle fuel assemblies that are currently used. The FHA radiological consequences are not increased because the fuel channel bow/bulge methodology used in the CSA does not impact the radiological source term of a single fuel assembly. Therefore, the proposed change does not significantly increase the probability or consequences of an FHA. Operation in accordance with the proposed change will not significantly increase the probability of a fuel mispositioning event because fuel movement will continue to be controlled by approved fuel handling procedures. These procedures continue to require identification of the initial and target locations for each fuel assembly that is moved. The consequences of a fuel mispositioning event are not changed because the reactivity analysis demonstrates that the new subcriticality criteria and requirements will be met for the worst-case fuel

mispositioning event. Operation in accordance with the proposed change will not change the probability of a seismic event. The consequences of a seismic event are not increased because the forcing functions for seismic excitation are not increased and because the mass of storage racks has not changed. Operation in accordance with the proposed change will not change the probability of a loss of SFP cooling event because the systems and events that could affect SFP cooling are unchanged. The consequences are not significantly increased because there are no changes in the SFP heat load or SFP cooling systems, structures or components due to the proposed change in fuel channel bow/bulge methodology used in the CSA.

ATTACHMENT 1 Evaluation of Proposed Change Page 7 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No Onsite storage of spent fuel assemblies in the QCNPS, Units 1 and 2, SFPs is a normal activity for which QCNPS has been designed and licensed. As part of assuring that this normal activity can be performed without endangering the public health and safety, the ability to safely accommodate different possible accidents in the spent fuel pool have been previously analyzed. These analyses address accidents such as radiological releases due to dropping a fuel assembly; and potential inadvertent criticality due to misloading a fuel assembly. The proposed change does not alter the method of fuel movement or spent fuel storage and does not create the potential for a new accident. The proposed use of a new fuel channel bow/bulge methodology for performing the QCNPS revised SFP CSA does not change or modify the fuel, fuel handling processes, spent fuel racks, number of fuel assemblies that may be stored in the pool, decay heat generation rate, or the SFP cooling and cleanup system. The limiting fuel assembly mispositioning event does not represent a new or different type of accident. The mispositioning of a fuel assembly within the fuel storage racks has always been possible. The proposed change involves a revised CSA for the QCNPS, Units 1 and 2, SFPs using a new fuel channel bow/bulge methodology. The associated analysis results show that the storage racks remain sub-critical, with substantial margin, following a worst-case fuel

misloading event. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No The proposed change involves a revised CSA for the QCNPS, Units 1 and 2, SFPs using a new fuel channel bow/bulge methodology. This change was evaluated for its effect on margins of safety related to criticality and spent fuel heat removal capability. QCNPS Technical Specifications Section 4.3, "Fuel Storage," Specification 4.3.1.1.a requires the spent fuel storage racks to maintain the effective neutron multiplication factor, keff, less than or equal to 0.95 when fully flooded with unborated water, which includes an allowance for uncertainties. Therefore, for SFP criticality considerations, the required safety margin is five percent.

ATTACHMENT 1 Evaluation of Proposed Change Page 8 The proposed change ensures, as verified by the associated criticality analysis, that k eff continues to be less than or equal to 0.95, thus preserving the required safety margin of five percent. The proposed use of a new fuel channel bow/bulge methodology for performing the QCNPS SFP CSA does not affect spent fuel heat generation or the spent fuel cooling systems. In addition, the radiological consequences of a dropped fuel assembly remain unchanged as the anticipated fuel damage due to a fuel handling accident is unaffected by the use of a new fuel channel bow/bulge methodology to perform the CSA. The proposed change also does not increase the capacity of the Unit 1 and Unit 2 spent fuel pools beyond the current capacity of no more than 3657 and 3897 fuel assemblies, respectively. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above evaluation, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, paragraph (c), and accordingly, a finding of no significant hazards consideration is justified.

4.3 Conclusions

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

5.0 ENVIRONMENTAL CONSIDERATION

EGC has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation." However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review,"

paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

ATTACHMENT 1 Evaluation of Proposed Change Page 9

6.0 REFERENCES

1. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "Request for License Amendment Regarding Transition to AREVA Fuel," dated February 6, 2015 2. Letter from B. Mozafari (U.S. NRC) to M. J. Pacilio (Exelon Generation Company, LLC), "Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments Regarding NETCO Inserts (TAC Nos. MF2489 and MF2490)(RS-13-148)," dated December 31, 2014 ATTACHMENT 4 AREVA, Inc. Affidavit

ATTACHMENT 5 AREVA, Inc. Report FS1-0024106, Revision 1.0, "Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version"

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 2/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page REVISIONS REVISION DATE EXPLANATORY NOTES 1.0 See 1 st page release date New document. This is the Nonproprietary version of FS1-0024092.

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 3/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page TABLE OF CONTENTS

1. PURPOSE .......................................................................................................................

................. 4

2. METHODOLOGY/APPROACH ........................................................................................................ 5

3. ASSUMPTIONS ...................................................................................................................

............ 6

3.1. ASSUMPTIONS

...................................................................................................................

.................... 6

3.2. ASSUMPTIONS

REQUIRING VERIFICATION ....................................................................................... 6

4. DETERMINATION OF VALUES ......................................................................................................

7 4.1. EQUIVALENT FAST FLUENCE .......................................................................................................

....... 7 4.2. EXPECTED FUEL ROD GROWTH ................

......................................................................................

... 7 4.3. CLADDING CREEP ................................................................................................................

................. 7

4.4. CHANNEL

BOW AND BULGE .........................................................................................................

....... 7 5. REFERENCES ....................................................................................................................

........... 13 LIST OF TABLES Table 1 Assembly Fast Fluence by Exposure ....................................................................................

.........

7 LIST OF FIGURES Figure 1 Mean Bulge by Exposure for Different Channel Geometries .........................................................

9 Figure 2 Bulge Standard Deviations by Exposure for Different Channel Geometries ...............................

10 Figure 3 Mean Bow by Exposure for Different Channel Geometries .........................................................

11 Figure 4 Bow Standard Deviations by Exposure for Different Channel Geometries ................................. 12

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 4/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page

1. PURPOSE As a part of the licensing of ATRIUM 10XM fuel assemblies for the Dresden and Quad Cities units, the customer, EXELON, has had spent fuel rack criticality analyses performed by another vendor. The submittal of these analyses to the NRC has resulted in two Requests for Additional Information (RAIs) related to AREVA fuel assembly and fuel channel performance. These RAIs read as follows: 11. Sections 2.3.11.1.1 and 2.3.11.1.2 of the HI-2146153 analysis explain that fuel rod growth, cladding creep, and crud buildup do not need to be evaluated because these factors are not expected to be significant at the peak reactivity burnup of the design basis lattice. Changes to the fuel rod geometry as a result of irradiation may result in a positive reactivity impact. Provide information regarding the expected fuel rod growth, cladding creep, and crud buildup at this burnup, and explain why the reactivity impact would not be significant.

13. Section 2.3.11.2 of the HI-2146153 analysis does not describe clearly how the geometry is changed to evaluate fuel channel bulging and bowing. The text refers to the channel outer exposed width tolerance, but it is not clear if the outer exposed width is changed by varying the channel inner width or the channel wall thickness.

Describe how the MCNP model was altered for Case 2.3.11.2.1. Also, discuss how the channel bowing tolerance was determined and how it bounds any expected ATRIUM 10XM channel bulging/bowing.

The customer has requested that such information be provided for fuel assembly exposures in the ranges of 10-20 GWd/MTU and 40-50 GWd/MTU. Specific AREVA information requested to support responses to these RAIs are: 1) expected fuel rod growth; 2) cladding creep; 3) crud buildup; and 4) channel bowing and bulge. For fuel rod crud buildup, such data are core-dependent and must be supplied by the customer.

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 5/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page

2. METHODOLOGY/APPROACH This Engineering Information Record provides values responsive to the requests described in §1 based on design bases related to fuel channel design or design basis data for the RODEX4 methodology, which is used to analyze the thermal-mechanical performance of ATRIUM 10XM fuel rods.

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 6/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page

3. ASSUMPTIONS 3.1. ASSUMPTIONS There are no assumptions made in this EIR. 3.2. ASSUMPTIONS REQ UIRING VERIFICATION There are no assumptions requiring verification as all values are obtained from quality records.

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 7/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page

4. DETERMINATION OF VALUES 4.1. EQUIVALENT FAST FLUENCE Since the growth of the Zircaloy comprising the fuel rods is well correlated with fast neutron fluence, mechanistic or correlative models of fuel rod growth and cladding creep are provided as a function of this fluence. Since the customer has requested the information described above as a function of assembly-averaged exposure, it is necessary to provide fast fluence ranges that correspond to those exposure ranges. Approximate values are shown in Table 1 [2, Fig. 2].

Table 1 Assembly Fast Fluence by Exposure

4.2. EXPECTED

FUEL ROD GROWTH For expected fuel rod growth, which includes both growth and PCMI-induced creep, appropriate values in are found in the response to RAI #3 in the first round of RAIs to the RODEX4 topical report [1, BAW-10247Q1(P), p. 4]. Upper limits for both of the exposure ranges, based on the BWR data compilation are

[ ]

for the 10 GWd/MTU to 20 GWd/MTU range, and

[ ] for the 40 GWd/MTU to 50 GWd/MTU range. 4.3. CLADDING CREEP Cladding creep, manifesting itself in diametral reduction is shown the RODEX4 topical report [1, Figure 4.16]. For the first exposure interval of 10 GWd/MTU to 20 GWd/MTU, the maximum reduction is less than

[ ], while for the second exposure interval of 40 GWd/MTU to 50 GWd/MTU, the maximum reduction is bounded by

[ ]. 4.4. CHANNEL BOW AND BULGE The RODEX4 methodology does not consider channel bulge as bulge occurs predominantly at axial locations where there is little or no in-channel voiding. In addition, the RODEX4 methodology places no limits on channel bowing, but explicitly computes bowing based on fast fluence gradients across the channel and accordingly penalizes linear power margins. As a part of the channel mechanical design methodology, channel bow and bulge statistics are provided as inputs to a Monte Carlo methodology for

assessing interference with control blades [3, §7.0]. Plots of these statistics are provided as function of exposure and lattice geometry in Figure 1 through Figure 4 [3, Figures 7.7 through 7.10]. The values denoted as D-lattice 2.54 mm correspond to 100 mil D-lattice channels, while those denoted as D-lattice 2.10 mm correspond to 80 mil D-lattice channels.

[

]

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 8/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page

[ ] reasonably bounds the observed population of channels for each exposure range.

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 9/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page Figure 1 Mean Bulge by Exposure for Different Channel Geometries

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 10/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page Figure 2 Bulge Standard Deviations by Exposure for Different Channel Geometries

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 11/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page Figure 3 Mean Bow by Exposure for Different Channel Geometries

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 12/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page Figure 4 Bow Standard Deviations by Exposure for Different Channel Geometries

N° FS1-0024106 Rev. 1.0 Information to Support EXELON RAI Responses for Spent Fuel Pool Criticality Analyses - Nonproprietary Version Handling: Restricted AREVAPage 13/13 AREVA - Fuel BL This document is subject to the restrictions set forth on the first or title page

5. REFERENCES 1 BAW-10247PA, Revision 0. Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors. April 2008.

2 [ ] 3 EMF-93-177(P)(A), Revision 1.

Mechanical Design for BWR Fuel Channels.

August 2005.

ATTACHMENT 6 Holtec International Affidavit

ATTACHMENT 7 Table 1: Criticality Safety Analysis (CSA) Approach Comparison, and Table 2: Comparison of the ATRIUM 10XM and SVEA-96 Optima2 Biases and Uncertainties (Nonproprietary Version)