ML20237F368

Amendment No. 11 to Materials License No. SNM-2506 for the Prairie Island Independent Spent Fuel Storage Installation Safety Evaluation Report

ML20237F368
Person / Time
Site:	Prairie Island
Issue date:	10/29/2020
From:	Office of Nuclear Material Safety and Safeguards
To:
WCAllen NMSS/DFM/STL 415.6877
Shared Package
ML20237F365	List: ML20237F366 ML20237F367 ML20237F368 ML20247J397
References
EPID L-2019LLA-0169
Download: ML20237F368 (19)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT INDEPENDENT SPENT FUEL STORAGE INSTALLATION MATERIALS LICENSE NO. SNM-2506 AMENDMENT NO. 11

SUMMARY

This safety evaluation report (SER) documents the review and evaluation of license amendment request (LAR) L-PI-19-009 (Agency Document Access Management System (ADAMS)

Accession No. ML19217A311) to Special Nuclear Material (SNM) License No. 2506 for the Prairie Island (PI) Independent Spent Fuel Storage Installation (ISFSI). By letter dated July 26, 2019, as supplemented April 29, 2020 and June 10, 2020, Northern States Power Company -

Minnesota (NSPM) submitted LAR L-PI-19-009 to the NRC in accordance with 10 CFR 72.56, to increase the storage capacity of the PI ISFSI and to approve the design of a new storage pad to be built at the existing facility.

The NRC staff evaluated the requested changes to assess whether the PI ISFSI continues to meet the applicable requirements of 10 CFR Part 72 for independent storage of spent fuel and of 10 CFR Part 20 for radiation protection. Staff followed the guidelines in NUREG-2215 Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities in its evaluation.

The NRC staff evaluated the proposed changes to SNM License No. 2506 in NSPMs amendment request. The evaluation did not reassess previously approved portions of the license, the TS, the Final Safety Analysis Report or those portions of the FSAR modified by NSPM, as allowed by 10 CFR 72.48, which are not associated with this amendment request.

The review disciplines evaluated are as described below for the requested change.

The staff reviewed the license amendment 11 request and the justifications for the requested changes. The NRC staff finds that the requested amendment to License No. SNM-2506 for the PI ISFSI meets the regulatory requirements of 10 CFR Part 72 based on the statements and representations in the application, as supplemented. In addition, editorial changes were made to the license.

1.0 GENERAL INFORMATION EVALUATION NSPM proposes to expand the PI ISFSI by constructing a third concrete storage pad parallel to the two existing concrete storage pads in a location approximately 38 feet (ft.) south of the existing eastern pad as shown in LAR Figure 2.2-1 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19217A311). NSPM stated that the new pad is designed to store 24 TN-40 or TN-40HT casks arranged in two parallel rows of twelve casks.

The NSPM design of the new pad has a nominal thickness of 36 inches (in.) to allow for the inclusion of duct banks, underground conduits, and pull boxes to monitor cask cover seal pressure to be consistent with the existing ISFSI pads. The new pad will also be the same length as the existing pads. However, the new pad will be 40 ft. wide versus the 36 ft. width of the existing pads. NSPM explained that the extra width will provide flexibility for the possible use of alternate cask technologies. However, NSPM clarified that cask spacing and centering on the new pad will be the same as for the existing pads. After reviewing the information presented, staff makes the following finding:

2 F1.1 The general description and discussion of the PI ISFSI presented in the LAR, with special attention to the design and operating characteristics, features, and principal considerations important to safety, are sufficient to familiarize a reviewer or stakeholder with the design.

The staff concludes that the general information presented in the LAR satisfies the general description requirements under 10 CFR Part 72. This finding is reached on the basis of a review that considered the regulation, applicable regulatory guides, and accepted practices.

2.0 SITE CHARACTERISTICS EVALUATION The requested changes do not impact or change previous evaluations of the site characteristics.

Therefore, a new evaluation was not required.

3.0 PRINCIPAL DESIGN CRITERIA EVALUATION NSPM proposed to increase the total mass of uranium stored at the PI ISFSI from 715.29 metric tons of enriched Uranium (TeU) to 1049.6 TeU. NSPM derived this 1049.6 TeU mass limit by assuming that the total number of casks to be stored at the PI ISFSI is 64. In addition, NSPM assumed that each of the 40 fuel assemblies loaded into the 64 casks is a Westinghouse Standard 14x14 fuel assembly which has the maximum allowable TeU per fuel assembly (i.e.,

0.41 TeU/assembly). After reviewing the information presented, staff makes the following finding:

F3.1 The docketed materials adequately identify and characterize the SNF to be stored at the PI ISFSI.

The NRC staff finds that the PI ISFSI design criteria and design bases were appropriately defined. The staff concludes that the principal design criteria for the PI ISFSI are acceptable and meet the regulatory requirements in 10 CFR Part 72. This finding is based on a review that evaluated the LAR against the NRCs regulations, the appropriate regulatory guides, the applicable codes and standards, and accepted engineering practices.

4.0 STRUCTURAL EVALUATION NSPMs LAR seeks to increase the maximum amount of spent fuel that may be possessed and stored at the PI ISFSI under SNM License No. 2506, as well as approval of the design of an additional concrete pad to be built within the confines of the existing ISFSI facility. NSPM stated that the proposed change does not impact or change the type or characteristics of either the cask technology or the spent fuel types authorized under SNM License No. 2506. In addition, NSPM stated that the new pads design and associated design functions are consistent with those of the two existing pads. NSPM used a different method of evaluation for the proposed pad than was in the PI ISFSI safety analysis report (SAR) for the existing pads. This chapter reviews the structural analysis of the proposed PI ISFSI concrete pad and the soil structure interaction (SSI) analysis to ensure consistency with the PI ISFSI design basis.

4.1 Prairie Island Storage Pad Description In LAR Section 2.2.2 (ADAMS Accession No. ML19210D273), NSPM stated the proposed concrete storage pad will be located within the existing ISFSI facility approximately 38 ft. south of the existing eastern pad. LAR Figure 2.2-1 provides the location of the new pad relative to

3 the current pads. NSPM explained that the new pad will have a nominal thickness of 3 ft. and a length of 216 ft. consistent with the design of the two current pads. The new pad width will be 40 feet versus 36 feet for the existing pads to allow for future flexibility and the possible use of alternate storage technologies. In response to request for additional information 2.3 (ADAMS Accession No. ML20162A445), NSPM stated that the new pad design is comparable to the existing pads. Specifically, #11 reinforcing bars are spaced at 12 in. intervals on the bottom face and #9 reinforcing bars are spaced at 12 in. intervals on the top face in both directions (longitudinal and lateral) for both faces, which is comparable to the existing pads. NSPM stated that the new pad design has the capacity to store either 24 TN-40 or 24 TN-40HT casks arranged in two parallel rows of twelve casks which is the same as the design of the existing pads.

The NRC staff determined NSPM sufficiently described pertinent structural details (e.g.,

dimensions, reinforcement and expected cask layout, etc.) for the proposed PI ISFSI storage pad. The staff also notes that the cask technology is the same as was approved for the current pads. For these reasons, the NRC staff finds that NSPM has adequately described the new ISFSI pad design.

4.2 New Storage Pad Structural Analysis 4.2.1 New Storage Pad Design Criteria and Loads NSPM discussed design inputs and criteria for the storage pads in LAR Section 3.4.1 (ADAMS Accession No. ML19210D273). NSPM stated that the new ISFSI pad design is based on NUREG-1567, NUREG-1536, ACI 318-08, ACI 349-13 and applicable portions of NUREG-0800.

NSPM used a lower bounding nominal compressive strength of 3000 psi for the storage pad structural analysis. NSPM considered the following key design inputs in analyzing the new storage pad:

• Bounding TN-40HT cask parameters

• Updated Cask Transport Vehicle (CTV) weight

• Static soil subgrade modulus was used for all load cases

• Enveloped accelerations from SSI analysis were conservatively increased from 0.125 for the existing pads to 0.25 g for both vertical and horizontal accelerations.

NSPM used criteria established in NUREG-1536 and ACI 349 to calculate governing load combinations for the pads; these calculations considered normal, off normal, postulated accidents, and natural phenomena. The calculated load combinations incorporate both strength and stability considerations. NSPM provided the governing load combinations in its LAR Section 3.4.2.2 (ADAMS Accession No. ML19210D273). The NRC staff finds that the design criteria and loads for the new storage pad are consistent with guidance in NUREG-1536 and ACI 349 for calculating governing load combinations for the pads and is therefore acceptable.

4.2.2 Analytical Approach 4.2.2.1 Model Characteristics In LAR Section 3.4.2.2 (ADAMS Accession No. ML19210D273) and NSPMs response to a request for additional information (ADAMS Accession No. ML20162A445), NSPM discussed various aspects of the pad structural analysis. NSPM considered the following model variables:

4 cask layout configurations, pad modeling and Cask Transport Vehicle (CTV) placement on the pad.

A total of four different cask load configurations were considered to envelope the worst-case moments and shear forces. The cask load configurations were as follows: single cask, 6 casks, 12 casks and 24 casks. In addition, NSPM examined the effects of the CTV placement locations in the four cask loading configurations to assess the impact the CTV had on the proposed ISFSI pad response. NSPM stated that CTV locations were chosen to maximize the pad shear and moment responses.

NSPM used the SAFE software for the structural analysis. NSPM modeled the storage pad using SAFE generated isotropic, rectangular, four-node thick-plate shell elements that combined membrane and plate-bending behavior. NSPM defined the shell element mesh to have maximum dimensions of 2 ft. x 2 ft. NSPM stated that the shell elements were used for modeling the concrete pad and that subgrade springs were used to model the supporting soil which is consistent with the analysis approach in SAR Section 4.2.1 for the existing pads.

The staff determined that the cask loading configurations used in the analysis are representative of bounding loading patterns within the pad. The staff finds that the model parameters are valid and appropriately characterize pad responses. The staff also finds that NSPM appropriately considered the effects of the CTV placements on the overall responses. Therefore, the staff finds that the model characteristics are representative of the proposed ISFSI pad.

4.2.2.2 Structural Capacity of the PI ISFSI Pad In LAR Section 3.4.2.3 (ADAMS Accession No. ML19210D273) and its response to a request for additional information (ADAMS Accession No. ML20162A445), NSPM discussed the design approach for the ISFSI pad. As discussed earlier, NSPM performed the pad structural analysis by analyzing various cask configurations and load combinations, including VCT loading, using SAFE. Based on the SAFE results, NSPM obtained maximum positive and negative moments in both the long and short directions of the pad. NSPM based the design mostly on shear and flexure by the design strips method. NSPM defined the design strips with an effective width of 13 ft. and centered the strips on the ISFSI casks. In accordance with the guidance in NUREG-1536, NSPM calculated the capacity over demand ratios to be greater than unity.

Based on the results of the analyses, NSPM concluded that a 3 ft. deep concrete pad having compressive strength of 3,000 psi reinforced with #11 reinforcing bars spaced at 12 in. intervals on the bottom face and with #9 reinforcing bars at 12 in. intervals on the top face in both directions (longitudinal and lateral) provides sufficient capacity to resist the maximum moment and shear forces. The NRC staff reviewed the overall approach and finds that NSPM used appropriately conservative criteria and applicable portions of ACI 349 and ACI 318 in its design calculations. The staff finds the structural capacity of the PI ISFSI storage pad is consistent with the guidance provided in NUREG-2215. For these reasons, the NRC finds the structural capacity of the PI ISFSI storage pad acceptable.

4.2.2.3 Storage Pad Stability In LAR Section 3.4.2.3 (ADAMS Accession No. ML19210D273) and its response to staffs request for additional information (ADAMS Accession No. ML20162A445), NSPM discussed soil bearing pressure for the storage pad. NSPM calculated a maximum bearing pressure of the pad against the soil as a function of the modulus of subgrade reaction and compared it to the

5 necessary soil bearing capacity. NSPM estimated a factor of safety greater than one. The staff finds this approach acceptable.

In responding to a request for additional information (ADAMS Accession No. ML20162A445),

NSPM discussed the storage pad settlement bearing pressure. NSPM presented values for settlement, differential settlement, and subgrade modulus for cask load configurations for a single cask, 6 casks, 12 casks, and 24 casks in Tables 2.15 through 2.19. NSPM concluded that pad stability is maintained for all cases. The NRC staff finds the subgrade modulus values that NSPM used show the ISFSI concrete pad to be flexible for settlement purposes, and that they demonstrate more conservative settlement when compared with a rigid pad. The staff finds that NSPMs settlement analysis is appropriately conservative because the small amount of expected bending moments due to small settlement will not adversely affect the structural capacity of the pad.

In LAR Section 3.2 (ADAMS Accession No. ML19210D273) and its response to a request for additional information (ADAMS Accession No. ML20162A445), NSPM discussed the new ISFSI pad soil liquefaction analysis. NSPM augmented previous site exploration performed as part of the analysis of the existing pads by conducting new Cone Penetration Tests. NSPM performed a numerical analysis to calculate a Factor of Safety against liquefaction using criteria in RG 1.198. NSPM stated that the site Factors of Safety were greater than 1.4 as recommended in RG 1.198. After reviewing the information provided, the NRC staff finds that NSPMs approach is consistent with the guidance of RG 1.198 and finds the soil liquefaction analysis acceptable.

4.3 Storage Pad Soil Structure Interaction (SSI) 4.3.1 Design Criteria related to SSI NSPM discussed design inputs for the SSI analysis in LAR Section 3.3.1 (ADAMS Accession No. ML19210D273). NSPM stated that the storage pad design is based on guidance presented in NUREG-0800, NUREG/CR-6865, RG 1.132, RG 1.61, and RG 1.92. NSPM considered the following key design inputs for the storage pad:

• Nominal pad dimensions,

• Average concrete compression strength of 3500 psi,

• Seed input accelerations from Pacific Earthquake Engineering Research,

• Safe Shutdown Earthquake ground motion of 0.12 g and 0.08 g in the horizontal and vertical directions, respectively, which is the design basis SSE ground motions for the existing pads, and

• No liquefaction was assumed per the results of the liquefaction analysis (see previous section).

4.3.2 Analytical Approach 4.3.2.1 Model Characteristics NSPM stated that it used artificial acceleration input motions (two horizontal, one vertical), which were developed following guidance in NUREG-0800. Figures 2.2 through figure 2.7 of NSPMs RAI response (ADAMS Accession No. ML20162A445) show the input accelerations and response spectra comparisons. NSPM stated that the generated acceleration input motions meet the recommended criteria for the strong motion duration, the absolute values of correlation

6 coefficient, the energy gap for the frequency range of interests, and the phase angle consistency in NUREG-0800. The staff finds that the input motions used for the SSI analysis are acceptable because they are consistent with the guidance in NUREG-0800.

Additionally, to account for potential effects of variability in the underlying rock and soil conditions, NSPM considered three different soil profiles. NSPM calculated best estimate, lower bound, and upper bound profiles based on field data results, in accordance with NUREG-0800.

NSPM calculated strain dependent soil properties for the underlaying soil beneath the pad through an iterative process using the SHAKE software. The staff finds that the calculation of strain dependent soil properties is consistent with the guidance of NUREG-0800; therefore, it is acceptable.

4.3.2.1 SSI Analysis NSPMs SSI approach consists of a structural model of the concrete pad with four node thick shell elements with interaction nodes between the pad and soil using SASSI 2010 software.

Because the TN-40HT cask is considered bounding in weight, NSPM modeled this cask as a vertical massless beam element with a lumped mass located at the cask center of gravity.

NSPM considered the four cask loading patterns described in SER Section 4.2. The staff notes that casks were modeled as rigid members and were configured to minimize any finite element stiffening effects on the concrete pad. The staff finds the finite element mesh used in modelling the pad is sufficiently refined to identify the pad bending moments between adjacent casks.

NSPM considered a total of 36 SSI analyses. These analyses represent multiple cask configurations, soil profiles, input motions, cracked pad conditions and uncracked pad conditions. NSPM concluded that cask responses are governed by the upper bound soil profile and that the maximum cask accelerations are 0.125 g and 0.128 g in the horizontal directions and 0.079 g in the vertical direction. The staff finds that NSPMs SSI analyses considered a wide variety of conditions that could affect the performance of both the pad and the cask. Staff also finds that the analyses were performed in a manner consistent with guidance in NUREG-0800. Therefore, the staff finds NSPMs analysis acceptable.

4.3.2.2 Cask Stability Analyses NSPM evaluated potential sliding and overturning of the casks using inequality checks with pertinent aspects of the cask and cask accelerations resulting from the SSI analysis. In performing the inequality checks, NSPM used an approach consistent with that identified in SAR Section A3.2.3.2. Using this approach, NSPM calculated factors of safety against sliding and overturning that are greater than one as described in SAR Section A3.2.3.2. The staff finds it is appropriate to use the cask accelerations generated from the SSI analyzes. In addition, the NRC staff finds the proposed amendment does not change the interaction between the cask and pad. Furthermore, the cask stability evaluations for the proposed pad are consistent with the previously approved method of evaluation. For these reasons, the staff also concludes that the cask stability analysis results are acceptable and provide reasonable assurance that the casks will not slide or overturn.

7

4.4 Evaluation Findings

F4.1 The SAR and docketed materials describe the design of the ISFSI structures in sufficient detail to support findings in 10 CFR 72.40, Issuance of License, for the term requested in the application, including the design criteria pursuant to Subpart F, the design bases, and the relation of the design to the design criteria, utilize applicable codes and standards, and therefore meet the requirements in 10 CFR 72.24(c)(1), 10 CFR 72.24(c)(2), and 10 CFR 72.24(c)(4) with respect to technical information.

F4.2 The SAR and docketed material contain information related to materials of the construction, general arrangement, dimensions of principal structures, and descriptions of all SSCs important to safety in sufficient detail to support a finding that the ISFSI will satisfy the design bases with an adequate margin of safety, and therefore meets the requirements in 10 CFR 72.24(c)(3) with respect to technical information.

F4.3 The licensee demonstrated that the proposed ISFSI pad is not susceptible to liquefaction and therefore meets the requirements of 10 CFR 72.103(c).

F4.4 The SAR and docketed materials adequately describe the design criteria for the SSCs important to safety and other SSCs, and therefore meet the requirements in 10 CFR 72.120(a).

F4.5 The SSCs important to safety are designed to withstand the normal and off-normal conditions associated with the site and can withstand postulated accidents, and therefore meet the requirements in 10 CFR 72.122(b)(1).

F4.6 The SSCs important to safety are designed to withstand the natural phenomena associated with the site without impairing their ability to perform their intended safety functions (with consideration for the most severe natural phenomena reported for the site and in the appropriate combination of normal and accident conditions), and therefore meet the requirements in 10 CFR 72.122(b)(2)(i).

The NRC staff concludes that the structural properties of the proposed PI ISFSI pad comply with the aforementioned regulations of 10 CFR Part 72 and that the applicable design and acceptance criteria are satisfied. The staffs review of the application considered appropriate regulatory guides, applicable codes and standards, and accepted engineering practices.

5.0 THERMAL EVALUATION This technical evaluation considers the thermal impact of the proposed expansion of the existing PI ISFSI, which currently contains two storage pads (west to east longitudinal array), each with a capacity of 24 TN-40 or TN-40HT casks. The proposed expansion would place an additional storage pad with a capacity of 24 casks (TN-40 or TN-40HT) south of the existing eastern pad.

NSPM specifically requested that staff 1) approve the construction and deployment of an additional ISFSI pad, and 2) approve an alternate method for to evaluate the heat transfer of the casks at the expanded ISFSI (limited to this LAR). NSPM provided technical details in support of the alternative method to demonstrate thermal evaluation safety which the staff analyzed below.

8 5.1 Evaluation Methodology 5.1.1 Existing Methodology The methodology used to analyze the current two-pad array at the ISFSI consists of two parts:

a detailed finite element analysis (FEA) model of the TN-40HT cask and an FEA-array evaluation. This methodology predicted increased cask surface temperatures due to the effects of the surrounding casks. NSPM also used this combined methodology to evaluate the temperatures of other cask components on the ISFSI pad (e.g., peak cladding and neutron resin temperatures). The results of the evaluation demonstrated significant margins when compared with component acceptance criteria for temperatures, with the exception of the neutron absorber, which had a 15°F margin for the existing longitudinal array. This indicated that the existing methodology was sufficiently robust to demonstrate a reasonable assurance of adequate protection for nearly all components that were important to safety, and that a new methodology would not likely challenge previous safety conclusions. The NRC staff subsequently focused on the alternate methodology to determine if the approach was representative of the thermal transfer mechanisms. The NRC staff also evaluated if the approach would produce significantly unconservative results for the neutron absorber temperatures whereby the acceptance criteria would not be met.

5.1.2 Proposed Alternate Methodology The proposed alternate methodology for determining the heat transfer effects of adding an additional storage pad consisted of the development of a simplified numerical model that was subsequently compared with the ANSYS FEA used for the existing storage pads. In order to determine the efficacy of this approach, NSPM executed a STAR CCM+ model of a single pad configuration with 24 casks and the results were compared with the previously evaluated ANSYS model used for the existing licensing basis (hereafter referred to as benchmark). The purpose for this benchmark was to demonstrate that the proposed alternate methodology for calculating external cask temperatures would produce the same results as the previously used and approved methodology.

NSPMs model included radiation heat transfer between casks, the environment, and the ground, but notably did not include convection around the casks. NSPM noted, that since the cask models themselves were simplified (i.e., they did not include the cask internal components) only the cask outer shell temperatures were considered valid results. Further, since NSPM did not report actual values for thermal conductivity to the ground, NSPM performed a sensitivity study for a range of values to determine the overall effects on the results.

NSPM reported an approximately 6°F temperature delta for the maximum cask temperature and an approximately 3°F temperature delta for the cask average temperature which shows that the results are essentially the same for both models. This demonstrates that the proposed alternate methodology produces reasonably similar results for a single pad configuration. The NRC staff noted that the results produced by the proposed alternate methodology are slightly unconservative (i.e., predicted lower temperatures compared to the ANSYS model) which is important when considering the reduced margin for the neutron absorber (i.e., the available margin could be 6°F less than the reported value of 15°F).

NSPM subsequently analyzed the effects of the proposed pad on the cask surface temperatures of the existing northern and eastern pad array. NSPM reported that the proposed pad would increase the maximum surface temperature of a cask by 2.8°F and increase the average

9 surface temperature of a cask by 1.5°F. NSPM noted that the radiation view factors determined for the existing evaluation of the TN-40HT are not appreciably affected by casks more than 74 ft. away. NSPM asserted that the casks on the proposed pad do not influence the existing calculated view factors since the distance to the nearest cask exceeds 74 ft. The results identified above support this assertion. In addition, as noted above, the alternate methodology/benchmark underpredicts the detailed/existing model results by 6°F for the cask maximum surface temperature and 3°F for the cask average surface temperature. This underprediction of temperature does not raise safety concerns because of the available margin of 15°F for the neutron absorber.

5.2 Confirmatory Evaluation The NRC staff performed a confirmatory analysis using the SINDA/FLUINT thermal code to determine whether the reported temperature values for the proposed ISFSI expansion provide reasonable assurance that an applicants results are adequate to make a safety determination.

The confirmatory analysis is not intended to exactly duplicate the surface temperatures identified in the LAR. The confirmatory analysis uses input values reported in the application and relies on assumptions based on engineering judgment. The assumptions, developed by the NRC staff include the percentage of radial heat flux, heat transfer being primarily radiation, and the effects of insolation.

The NRC staffs model analyzed the existing fully loaded arrays on the northeast pad and a fully loaded proposed array on the southern pad. The NRC staff calculated radiation view factors, for the fully loaded northeast pad and the fully loaded southern pad, that were consistent (slightly greater) with those reported by NSPM. The NRC staffs analysis predicted cask surface temperatures approximately 11°F higher than the temperature predicted by NSPM which is expected given the higher view factors.

5.3 Conclusion The thermal evaluation submitted by the applicant and the results of the staffs confirmatory analysis are acceptable to the NRC staff because they illustrate negligible temperature increases to important to safety components which demonstrates that the ISFSI expansion will have a minimal thermal performance safety impact at the PI site.

5.4 Findings

F5.1 SSCs important to safety are described in sufficient detail in the SAR to enable an evaluation of the effectiveness of their heat removal capability, as required by 10 CFR 72.122(b), 10 CFR 72.122(f), 10 CFR 72.122(h), 10 CFR 72.122(i), and 10 CFR 72.122(j) 10 CFR 72.128(a)(4). Storage container structures, systems, and components important to safety remain within their operating temperature ranges as required by 10 CFR 72.92(a), 10 CFR 72.120(a), 10 CFR 120(d), and 10 CFR 72.128(a).

F5.2 The SNF cladding is protected against degradation that could lead to gross ruptures under off-normal and accident conditions by maintaining cladding temperature below 400°C (752°F) in an inert environment. Protection of the cladding against degradation is expected to allow ready retrieval of the SNF for further processing or disposal in as required by 10 CFR 72.122(h).

10 The NRC staff concludes the thermal design of the PI ISFSI is in compliance with 10 CFR Part 72 and the applicable design and acceptance criteria identified in the SAR are satisfied. The evaluation of the thermal design provides the NRC staff with reasonable assurance that PI ISFSI will allow safe storage of SNF. The NRC staff concludes that the thermal evaluation of the proposed LAR meets the requirements of 10 CFR Part 72, and follows the appropriate regulatory guides, applicable codes and standards, and accepted engineering practices.

6.0 SHIELDING EVALUATION The requested changes do not impact or change any previous shielding evaluation. Therefore, an evaluation was not required.

7.0 CRITICALITY EVALUATION

The requested changes do not impact or change any previous criticality evaluation. Therefore, an evaluation was not required.

8.0 MATERIALS EVALUATION The NRC staff conducted the materials review of the PI ISFSI license amendment to verify that the materials of the structures, systems, and components (SSCs) will perform adequately with the expanded fuel storage capacity. The materials review focused specifically on the concrete design code for the new ISFSI pad, the potential effects of increased cask component temperatures due to the additional cask inventory, and the implementation of aging management activities to the additional casks and pad.

8.1 Codes and Standards The existing ISFSI pad design was designed to meet the criteria of ACI 349-85 (including the 1990 supplement), Code Requirements for Nuclear Safety-Related Concrete Structures, which specifies a concrete compressive strength range of 3,000-4,000 psi. In SAR Section A4.2.1, NSPM stated that the proposed additional concrete pad is designed to the criteria of the more recent 2013 edition of the ACI code. Because ACI 349-13 specifies a minimum concrete compressive strength of 4,500 psi, NSPM proposed to take an exception to this criterion to be consistent with the 3,000-4,000 psi strength of the existing ISFSI pad design.

NSPM noted that the 4,500 psi minimum strength criterion in ACI 349-13 is present to ensure that the concrete has adequate durability when exposed to freezing and thawing cycles. To ensure concrete durability at the lower 3,000-4,000 psi strength, NSPM stated the concrete for the proposed pad will be fabricated following the guidelines in ACI 201.2R-16, Guide to Durable Concrete, for severe (Exposure Class F2) environments, which recommends minimizing damage due to weathering, chemical attack, abrasion, and other deterioration processes through appropriate water-to-cement ratios, supplementary cementitious materials, and good workmanship.

The staff reviewed the concrete pad design and verified that, except for the strength criterion, it is consistent with ACI 349-13. The staff finds using the 2013 edition of ACI 349 for the construction of the additional ISFSI pad is consistent with the recommendations in NUREG-2215. The staff also finds the exception to the strength criterion of the 2013 edition of the ACI code is acceptable because NSPM proposed to follow the concrete construction practices in ACI 201.2R. The staff concludes that the durability performance of the concrete can be

11 adequately achieved by following the ACI 201.2R guidelines and using operating experience insights gained from the ongoing aging management inspections of the existing pads, which were also designed to a 3,000-4,000 psi strength. NSPMs renewed ISFSI license requires visual inspections of the existing pad be performed using the intervals and acceptance standards in ACI 349.3R-96, "Evaluation of Existing Nuclear Safety-Related Concrete Structures." In addition, as described in SER Section 8.5, after the new pad reaches 20 years of service, inspections under the aging management program are required to verify that freezing and thawing cycles are not causing aging degradation that could challenge the pads performance.

8.2 Radiation Shielding Materials In SAR Section A3.3.2.2.4.1.1 and Table A3.3-3, NSPM stated that placement of the additional storage casks at the ISFSI would increase the maximum temperature of the neutron shielding polyester resin by 5°F (3°C). As a result, the maximum temperatures of the top and radial neutron shields will increase to 196°F (91°C) and 290°F (143°C), respectively, for normal and off-normal conditions of storage. In the fire accident, NSPM assumed the resin decomposed and, therefore, is not credited for shielding.

The staff noted that the temperatures of the radiation shielding resin will remain below the 300°F (149°C) allowable temperature established in the SAR for previously approved license amendments. Also, shielding performance of the existing casks must be verified by periodic dose rate surveys of the casks in accordance with the sites aging management program in its renewed license. These surveys will also be performed on the new casks when they reach twenty years of service. As a result, the staff finds that the shielding performance of the materials remains acceptable.

8.3 Seals In SAR Section A3.3.2.2.4.1.1 and Table A3.3-3, NSPM stated that placement of the additional storage casks at the ISFSI would increase the maximum temperature of confinement seals by 5°F (3°C). As a result, the maximum temperatures of the lid seal, vent seals, and port seals increased to 189°F (87.2°C), 190°F (87.8°C), and 190°F (87.8°C) respectively, during normal and off-normal conditions of storage. In a fire accident, the maximum temperatures of the lid seal, vent seals, and port seals increased to 379°F (193°C), 270°F (132°C), and 270°F (132°C) respectively.

The NRC staff finds that the temperatures of the seals will remain below 536°F (280°C), the maximum allowable temperature established in the SAR for previously approved license amendments, with the expansion of the spent fuel inventory at the ISFSI. For this reason, the staff concludes that the seal performance remains acceptable.

8.4 Spent Fuel Cladding In SAR Section A3.3.2.2.4.1.1 and Table A3.3-3, NSPM stated that placement of the additional storage casks at the ISFSI would increase the maximum temperature of the spent fuel cladding by 2°F (1°C). As a result, the maximum temperatures of the cladding increased to 696°F (369°C) during normal and off-normal conditions of storage, to 790°F (421°C) in a fire accident, and to 909°F (487°C) in a buried cask accident.

12 The staff finds the temperatures of the spent fuel cladding with the expansion of the spent fuel inventory at the ISFSI will remain below the allowable temperature limits established in the SAR for previously approved license amendments (752°F [400°C] for normal and off-normal conditions and 1058°F [570°C] in an accident). For this reason, the staff concludes that the fuel cladding performance remains acceptable.

8.5 Management of Aging Degradation The PI ISFSI license was renewed for an additional 40 years on December 9, 2015, and the site currently is implementing the aging management programs (AMPs) that were conditions of the renewed license. NSPM stated that the additional casks and new pad associated with the amendment do not involve any modification to the types of spent fuel or cask technology.

Therefore, NSPM concluded that the existing AMPs and time-limited aging analyses (TLAAs) under the sites renewed license are equally applicable to the additional SSCs. In accordance with the guidance in NUREG-2215, the staff reviewed the amendment application and license renewal documentation (NSPM, 2011 and 2015) to ensure that NSPM appropriately evaluated the need to revise (1) the scope of SSCs subject to aging management, (2) the aging management review for applicable aging mechanisms and effects, and (3) the TLAAs and AMPs to ensure that they remain effective to manage the aging of all SSCs.

8.5.1 Scope of SSCs requiring aging management NSPM stated that the amendment does not involve any changes to the design of the site SSCs.

As a result, NSPM did not revise the scoping evaluation that was performed for the renewal of the PI ISFSI license. The staff reviewed the current amendment request and confirmed that the amendment does not add or remove any SSC designs to the storage system or change the safety classification of any SSCs. As a result, the staff finds that the existing scope of components addressed by the aging management activities remains valid and that NSPMs aging management scoping determination is acceptable.

8.5.2 Aging management review NSPM did not revise the credible aging mechanisms and effects that had been identified during the renewal of the PI ISFSI license in 2015. The NRC staff reviewed the amendment request and confirmed that the current amendment does not introduce any operating environment changes that could impact the aging management review that was performed for the license renewal. As documented above in SER Sections 8.2, 8.3, and 8.4, the additional spent fuel inventory associated with the amendment would cause small increases (5°F [3°C] or less) in the maximum temperature for some components. However, the staff finds that the component temperatures remain within allowable limits and that prolonged exposures to these temperatures do not introduce any new aging mechanisms and effects that were not previously addressed in the license renewal. As a result, the staff concludes that the aging management review that was performed for the license renewal remains valid for the new SSCs.

8.5.3 Time-Limited Aging Analyses NSPM did not revise the TLAAs that were originally performed for the renewal of the PI ISFSI license. The TLAAs include (1) the effect of elevated temperature on the creep behavior of aluminum basket components and (2) the effect of neutron radiation damage on metallic cask components. The staff reviewed the TLAAs to verify that they remain valid for the proposed amendment.

13 For the aluminum creep TLAA, the SAR Section A4B.1.5.6 analysis evaluated creep behavior at an operating temperature of 470°F (243°C) to demonstrate that the plastic strain would not reach the allowable threshold of 0.01 in 550,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (62.8 years). As shown in SAR Table A3.3-3, the current amendment increases the maximum normal and off-normal temperatures of the aluminum basket rails to 471°F (244°C). The NRC staff reviewed the potential effects of increasing the rail temperature by 1°F. The original TLAA demonstrated that the aluminum creep was below the allowable strain threshold by a very large margin, and the NRC staff determined the impact of the small increase in the basket rail temperature to be negligible.

Therefore, the NRC staff concludes that the original TLAA remains valid.

Regarding the radiation damage TLAA analysis, SAR Section A4.2.3.5 used an estimate of the neutron flux inside a TN-40HT cask, which is approximately three times greater than the neutron flux inside a TN-40 cask. NSPM concluded that the integrated neutron flux over 60 years would be about three orders of magnitude below levels that would degrade the performance of the metallic cask materials. Since the neutron flux inside the TN-40HT cask is unchanged by the amendment, NSPMs original TLAA conclusion that neutron radiation will not degrade the metallic structural material remains applicable. Therefore, the staff finds that the original TLAA remains valid.

8.5.4 Aging Management Programs NSPM stated that the existing license renewal AMPs remain applicable to the amended license.

The staff reviewed NSPMs evaluation of the applicability of the renewal AMPs and independently confirmed that the amendment neither changes the existing scope of components addressed by the aging management activities nor introduces changes to the component materials, geometries, or operating conditions that could affect the aging mechanisms and effects requiring management. Therefore, the NRC staff finds the evaluation to be acceptable.

8.5.5 AMP Implementation Timing NSPM also proposed to implement the AMPs for each SSC added during the renewed period of operation (i.e., after December 9, 2015) when those SSCs reach 20 years of service (ADAMS Accession No. ML20162A445). The staff noted that this approach differs from the current ISFSI license requirements. Under the existing renewed license, all SSCs are managed by the approved aging management programs, regardless of the time the SSCs have been in service.

In its evaluation of NSPMs proposal, the NRC staff noted that the original PI ISFSI license was approved for a 20-year term without AMPs. The AMPs were added to the renewed license based on an assessment of credible aging effects for SSCs in service longer than 20 years.

Although AMPs were not required prior to the license renewal, the original license required 1) quarterly visual inspections to verify that there is no significant damage to the casks, and 2) quarterly readings of the thermoluminescent dosimeters (TLDs) at the ISFSI site fence to confirm that area dose rate limits are not exceeded. These surveillance requirements remain in the ISFSI license and would apply immediately to casks added as a result of this amendment.

The staff finds NSPMs proposal to begin aging management activities when newly added SSCs reach 20 years acceptable because the approach is consistent with the original issuance of the ISFSI license for the first 20 years (i.e., performing visual inspections and TLD readings, but having no AMP requirements). The staff notes that the additional casks will be subject

14 immediately to the visual inspection and radiation monitoring surveillance requirements of the ISFSI license. Furthermore, these activities are capable of identifying any significant degradation of the cask materials during the first 20 years of service. The NRC staff added a 20-year implementation schedule to the existing License Condition Nos. 22, 23, and 24 to monitor the concrete pad, casks, and polymer-based neutron shields (i.e., SSCs) during the period of extended operation (ADAMS Accession No. ML20254A114).

8.6 Findings and Conclusions F8.1 NSPM has met the requirements in 10 CFR 72.24(d) and 10 CFR 72.128(a). The properties of the materials in the storage facility design have been demonstrated to support the safe storage and handling of spent nuclear fuel for the storage term under normal, off-normal, and accident conditions.

F8.2 NSPM has met the requirements in 10 CFR 72.124(b). Neutron absorbing materials are demonstrated to effectively control criticality without significant degradation over the storage life.

F8.3 NSPM has met the requirements in 10 CFR 72.120(d), 10 CFR 72.122(b)(1), and 10 CFR 72.124(b). Materials and storage contents are compatible with their operating environment such that there will be no adverse degradation or significant chemical or other reactions.

F8.4 NSPM has met the requirements in 10 CFR 72.122(h)(1). The spent fuel cladding has been demonstrated to be adequately protected against gross ruptures, or the fuel has been demonstrated to be otherwise confined.

F8.5 NSPM has met the requirements in 10 CFR 72.24(c)(4) and 10 CFR 72.122(a). The use of codes and standards, quality assurance programs, and control of special processes are demonstrated to be adequate to ensure that the design, testing, fabrication, and maintenance of materials support SSC intended functions.

The NRC staff reviewed the application under the applicable regulations, standard review plans, appropriate regulatory guides, applicable codes and standards, and accepted engineering practices. Based on its review, the NRC staff finds that the PI ISFSI design adequately considers material properties, environmental degradation and other reactions, fuel clad integrity, and material quality controls such that the design is in compliance with 10 CFR Part 72. The NRC staff concludes that there is reasonable assurance that the PI ISFSI will ensure the safe storage of spent fuel.

References

1. ACI 201.2R-16, Guide to Durable Concrete, American Concrete Institute, Farmington Hills, MI. November 2016.

2. ACI 349-85, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, with 1990 Supplement. American Concrete Institute, Farmington Hills, MI, March 1, 1990.

3. ACI 349-13, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, American Concrete Institute, Farmington Hills, MI, Errata as of June 2016.

4. ACI 349.3R-96, "Evaluation of Existing Nuclear Safety-Related Concrete Structures,"

American Concrete Institute, Farmington Hills, MI, March 1996.

15

5. NSPM, 2011. Prairie Island Independent Spent Fuel Storage Installation Application for Renewed ISFSI Site-Specific License, Northern States Power Company, October 20, 2011.

ADAMS Accession No. ML113140518

6. NSPM, 2015. Supplement to Prairie Island Independent Spent Fuel Storage Installation License Renewal Application - Revised Aging Management Plan, Northern States Power Company, October 12, 2015. ADAMS Accession No. ML15285A007.

9.0 CONFINEMENT EVALUATION The requested changes did not impact or change any previous confinement evaluation.

Therefore, an evaluation was not required.

10.0 RADIATION PROTECTION EVALUATION The radiation protection evaluation seeks to ensure that the proposed amendment 11 request submitted by NSPM fulfills the following acceptance criteria:

• The SAR demonstrates that the PI ISFSI meets the requirements in 10 CFR 72.104, Criteria for radioactive materials in effluents and direct radiation from an ISFSI or MRS, 10 CFR 72.106, Controlled area of an ISFSI or MRS, and 10 CFR 20.1301(a)(1),

Dose limits for individual members of the public, regarding annual doses.

• Dose rates, design features, and operations for the facility are consistent with and demonstrate appropriate consideration for As Low As Reasonably Achievable (ALARA) principles and objectives.

NSPM requested a license amendment to renewed SNM License No. 2506 to increase the maximum amount of spent fuel to 1,049.60 TeU. NSPM also requested the approval of the design of an additional concrete pad to be built within the PI ISFSI utilizing methods described in the existing PI ISFSI SAR.

10.1 Radiation Protection Design Features The design features associated with the PI ISFSI include the cask technology, an engineered radiation barrier (i.e., a berm), the type of fuel stored at the ISFSI, the amount of fuel stored at the ISFSI and the location of the fuel stored at the ISFSI. NSPM proposed increasing the amount of fuel stored at the ISFSI and adding a new fuel storage location (i.e., a new pad).

Therefore, staff reviewed only these design changes and their impact on the ISFSI.

NSPM designed the new pad to store either 24 TN-40 or TN-40HT casks arranged in two parallel rows of twelve casks which is consistent with the design of the existing pads. NSPM said that they intend to locate the new pad within the ISFSI facility approximately 38 ft. south of the existing eastern pad. NSPM also planned to increase the amount of material authorized for storage from 715.29 TeU to 1049.6 TeU. NSPM calculated the increase in special nuclear material using the uranium mass of the 14x14 Westinghouse fuel assembly with fuel inserts which is the design basis fuel assembly for the PI ISFSI. The NRC staff finds this assumption conservative because the14x14 Westinghouse standard fuel assembly with fuel inserts has the greatest uranium mass per fuel assembly versus the other fuel assemblies authorized for storage at the PI ISFSI. In addition, it has the strongest radiological source term. Based on the information provided, staff finds NSPM adequately described the ISFSI design features.

16 10.2 MCNP Methodology, Model and Assumptions NSPM utilized the Monte Carlo N-Particle (MCNP) computer code for all shielding calculations.

For neutrons, the MCNP computer code accounted for all reactions using the cross section set.

For photons, the code accounted for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, absorption in pair production with local emission of annihilation radiation, and bremsstrahlung. The NRC staff notes that the MCNP computer code, due to its capabilities, has wide-ranging applications and uses within the nuclear industry; this ensures that the MCNP computer code is well-vetted. In addition, the MCNP computer code is accepted by the NRC for these types of applications. For these reasons, NRC staff finds the MCNP computer code acceptable for use in the present application.

NSPM evaluated the off-site dose rates by modeling the new ISFSI pad with one 2x8 array of TN-40HT casks. NSPM used TN-40HT casks in the model to bound ISFSI operations with a mixture of TN-40 and TN-40HT casks. NSPM modeled the new pad and additional casks at the center of the existing ISFSI facility, between the two existing ISFSI pads location, versus its actual location 38 ft. south of the existing eastern ISFSI pad. NSPM also reduced the berm height around the ISFSI in the MCNP input model from 20.5 feet to 17 feet to accurately reflect the actual configuration. In developing the radiological source terms for the MCNP model, NSPM assumed that four casks were placed on the ISFSI pad every two years, and that each cask was loaded with the design basis fuel at the time of initial loading. For normal and off-normal conditions, NSPM modeled intact TN-40HT casks. For accident conditions, NSPM assumed the neutron shield and steel outer shell of the TN-40HT casks were not present which bounded the accident conditions described in SAR Section A8. NSPM employed variance reduction techniques in their MCNP model to track particles through very complex deep penetration situations and to minimize the computing expense needed to obtain a result. NSPM utilized mesh tallies to calculate average dose rates on the side, top and bottom TN-40HT cask surfaces. NSPM also calculated maximum dose rates above and below the radial neutron shield. NSPM used F5 tallies (i.e., point detectors) to calculate dose rates at distances away from the cask. NSPM positioned these detectors along the ISFSI sides and in the ISFSI corner locations to determine the dose rates as a result of the source symmetry. NSPM provided these detector locations in SAR Table A7A.7-1.

The NRC staff finds that the location of the new ISFSI pad in the MCNP model is conservative because it will produce higher predicted exposure rates versus the exposure rates from actual operations. NRC staff finds that using only TN-40HT casks in the MCNP model is also conservative since the TN-40HT cask is designed to store a stronger radiological source term than the TN-40. NRC staff finds the berm height used in the MCNP model reasonable because it reflects the actual ISFSI berm height. NRC staff finds the radiological source term assumption reasonable because it accounts for the actual radiological source term decrease over time.

NRC staff finds the MCNP tally types employed by NSPM to calculate dose rates are appropriate. The NRC staff finds the MCNP model employed by NSPM is reasonable based upon the reasons given above.

10.3 Offsite Collective Dose Assessment In its calculation of the collective offsite dose, NSPM conservatively assumed that the entire permanent population within a 2-mile radius of the plant was located at the residence subject to the highest exposure (0.45 miles northwest of the ISFSI). NSPM also calculated a collective dose for the large transient population of persons who are either employed at or visit the

17 Treasure Island Hotel and Casino which is located within a 2-mile radius of the plant. In making these calculations, NSPM conservatively assumed that this entire transient population is located 0.8 miles from the ISFSI. NSPM used the estimates of both permanent and transient populations within a 2-mile radius in the Prairie Island Nuclear Generating Plants evacuation time study (ADAMS Accession No. ML12363A173). A description of the off-site locations considered in the current evaluations, the relevant population data, distances and occupancy times are shown in SAR Table A7.5-2. NSPM identified that the maximum annual dose rate to the nearest resident is 4.34 mrem/yr. NSPM provided this maximum annual dose rate in SAR Section A7A.7.2. NSPM stated that the existing ISFSI capacity, when fully loaded with 48 casks, would contribute 3.05 mrem/yr., and that the new pad, when fully loaded with sixteen casks, would contribute 1.27 mrem/yr. Based on the 2017 Annual Radioactive Effluent Report and Offsite Dose Calculation Manual, (ADAMS Accession No. ML18134A310), NSPM estimated that planned plant discharges would contribute an additional 1.50 x 10-2 mrem/yr.

The NRC staff finds that the maximum annual dose rate of 4.34 mrem/yr estimated by NSPM is below the radiation exposure limit of 25 mrem/yr to any real individual located beyond the controlled area boundary identified in 10 CFR 72.104. The NRC staff also finds the maximum annual dose rate of 4.34 mrem/yr is also well below the radiation exposure limit of 100 mrem/yr for individual members of the public provided in 10 CFR 20.1301(a)(1). NSPM identified that the maximum annual dose rate to the off-site population within a two-mile radius of the PI ISFSI is 0.45 mrem/hr. NSPM provided this maximum annual dose rate in SAR Section A7A.7.2.

NSPM stated that the existing ISFSI capacity, when fully loaded with 48 casks, would contribute 0.313 mrem/hr., and that the new pad, when fully loaded with sixteen casks, would contribute 0.133 mrem/hr. Based on their existing analyses, NSPM estimated that the radiation level at the ISFSI controlled area boundary from sources other than the spent fuel storage casks would be insignificant. The NRC finds this maximum dose rate is within the regulatory limit of 2 mrem/hr. provided in 10 CFR 20.1301(a)(2).

The NRC staff finds that the revised dose rates at the site boundary and for the nearest resident continue to meet the regulatory limits of 10 CFR 72.104(a), and 10 CFR 20.1301(a). The NRC staff concludes that the addition of 16 TN-40HT casks to the PI ISFSI will result in radiation levels at the site boundary and at the nearest resident that are within regulatory limits and therefore acceptable.

10.4 As Low As Reasonably Achievable As stated earlier, NSPM proposed no changes to the types of spent fuel or cask technology authorized for storage at the PI ISFSI as well as no changes to the berm surrounding the ISFSI.

In addition, NSPM made no changes to the ISFSI operating procedures used to ensure that occupational exposures remain ALARA. However, the proposed change involves increasing the quantity of spent fuel stored at the PI ISFSI, and for this reason NSPM analyzed the radiological impact from storing an additional sixteen TN-40HT casks at the ISFSI to ensure occupational exposures remain ALARA. NSPM identified fourteen personnel locations and updated the dose rates for these locations in SAR Table A7.4-3. The proposed ISFSI expansion increased the radiological worker collective dose, from a total exposure of 12.88 person-rem to 20.5 person-rem as shown in SAR Table A7.4-4. NSPM acknowledged that actual loading and maintenance activities may result in deviations from dose rates reported in the tables. In addition, localized regions of elevated dose rates at the ISFSI site are anticipated. However, NSPM remained committed to minimizing personnel exposures through good ALARA practices such as limiting time in radiation areas. Consistent with previous dose analyses, NSPM analyzed the ISFSI expansion utilizing the TN-40HT cask and the bounding design basis source terms. NSPM

18 presented the expected dose rates (for normal, off-normal, and accident conditions) for the TN-40HT cask in SAR Table A7A.2-1. SAR Figure A7A.2-1 illustrated the dose point locations.

SAR Table A7A.2-2 presented the total (direct plus skyshine) dose rates as a function of distance from the cask. The NRC staff finds that these reported dose rates demonstrate reasonable assurance that occupational exposure is ALARA. The NRC staff also finds that NSPMs analysis provides reasonable assurance that occupational exposure will be maintained ALARA.

10.5 Evaluation Findings F10.1 The SAR includes adequately detailed descriptions of the design and operational characteristics of SSCs in the PI ISFSI, including design criteria and design bases for the radiation protection evaluation and the radioactive materials to be stored at the facility, in compliance with 10 CFR 72.24(b), 10 CFR 72.24(c), 10 CFR 72.24(l), 10 CFR 72.120(a), 10 CFR 72.120(b), and 10 CFR 72.120(c). The SAR also includes evaluations of the performance of the facilitys SSCs important to safety with respect to radiation protection, in compliance with 10 CFR 72.24(d).

F10.2 The PI ISFSI is on the same site as Prairie Island Nuclear Generating Plant, Units 1 and

2. The cumulative effects of the combined operations of these facilities will not constitute an unreasonable risk to the health and safety of the public, in compliance with 10 CFR 72.122(e).

F10.3 The SAR provides analyses showing that the cumulative effects of the combined operations of these facilities will be within the dose limits given in 10 CFR 72.104(a).

These analyses include both direct radiation and effluent releases from Prairie Island Nuclear Generating Plant, Units 1 and 2 to the general environment during normal operations and anticipated occurrences. The SAR also includes appropriate and adequate operational restrictions and limits to meet the limits in 10 CFR 72.104(a) and ALARA objectives in compliance with 10 CFR 72.104(b) and 10 CFR 72.104(c).

F10.4 The SAR provides analyses that show that the doses to members of the public will not exceed the limits in 10 CFR Part 20.

The NRC staff finds reasonable assurance that the radiation protection design and program for the PI ISFSI meet the requirements in 10 CFR Part 20 and 10 CFR Part 72 and that the applicable design and acceptance criteria are satisfied. The NRC staff also finds reasonable assurance that the facility design, operations, and programs are adequate to ensure compliance with the regulatory dose limits and ALARA requirements in 10 CFR Part 20 and 10 CFR Part 72 for personnel and the public. The NRC staffs evaluation of the radiation protection program, facility design features, and ALARA objectives provides reasonable assurance that the PI ISFSI will provide for the safe storage of SNF. The NRC staff reached this finding after consideration of all applicable NRC regulations and regulatory guides, applicable codes and standards, accepted health physics practices, the statements and representations contained in the SAR, and the staffs confirmatory analyses.

11.0 OPERATION PROCEDURES AND SYSTEMS EVALUATION The requested changes do not impact or change any previous operation procedures and systems evaluation. Therefore, an evaluation was not required.

19 12.0 CONDUCT OF OPERATIONS EVALUATION The requested changes do not impact or change any previous conduct of operations evaluation.

Therefore, an evaluation was not required.

13.0 WASTE MANAGEMENT EVALUATION The requested changes do not impact or change any previous waste management evaluation.

Therefore, an evaluation was not required.

14.0 DECOMMISSIONING EVALUATION The requested changes do not impact or change any previous decommissioning evaluation.

Therefore, an evaluation was not required.

15.0 QUALITY ASSURANCE EVALUATION The requested changes do not impact or change any previous quality assurance evaluation.

Therefore, an evaluation was not required.

16.0 ACCIDENT ANALYSIS EVALUATION See SER Sections 4, 5, 8 and 10 for the NRC staffs accident evaluation of the LAR.

17.0 TECHNICAL SPECIFICATIONS EVALUATION The requested changes do not impact any previous technical specification evaluation. Therefore, an evaluation was not required.

ENVIRONMENTAL REVIEW Pursuant to 10 CFR Part 51, an environmental assessment (ADAMS Accession Number ML20275A342) has been prepared for this action and a finding of no significant impact (FONSI) was issued. The EA and FONSI were published in the Federal Register on October 8, 2020 (85 FR 63588).

CONCLUSION The NRC staff reviewed the license amendment request for SNM-2506, as supplemented, including the engineering analyzes, the proposed FSAR revisions, and other supporting documents. Based on the information provided in the application, as supplemented, the NRC staff concludes that SNM-2506, as amended, meets the requirements of 10 CFR Part 72.

Renewed SNM License No. 2506, Amendment No. 11, on October 29, 2020.