ML24263A206

Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 for LaSalle County Station - Holtec MPC-68MCBS

ML24263A206
Person / Time
Site:	LaSalle, Holtec
Issue date:	09/19/2024
From:	Para W Constellation Energy Generation
To:	Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation, Document Control Desk
References
RS-24-092
Download: ML24263A206 (1)

Text

200 Energy Way Kennett Square, PA 19348 www.constellation.com Constellation

10 CFR 72.7 RS-24-092

September 19, 2024

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

LaSalle County Station, Unit 1 and 2 Renewed Facility Operating Licensee Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373, 50-374, and 72-70

Subject:

Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 for LaSalle County Station - Holtec MPC-68MCBS

Pursuant to 10 CFR 72.7, Specific Exemptions, Constellation Energy Generation, LLC (CEG) requests an exemption from the requirements of 10 CFR 72.212(a)(2), 10 CFR 72.212(b)(3), 10 CFR 72.212(b)(5)(i), 10 CFR 72.212(b)(11), and 10 CFR 72.214 for the LaSalle County Stations Independent Spent Fuel Storage Installation (ISFSI). Specifically, an exemption is requested for the Holtec 68 Multi-Purpose Canisters (MPC) with a Continuous Basket Shim (MPC-68MCBS) design basis condition requiring analysis of a postulated non-mechanistic tip-over event.

The requested exemption will allow continued storage of loaded storage casks with MPC-68MCBS canisters, as listed in Table 1. Additionally, the exemption will allow future loading of MPC-68MCBS canisters, as listed in Table 2.

The exemption is needed because although Holtec performed a non-mechanistic tip-over analysis with favorable results and subsequently implemented the Continuous Basket Shim (CBS) design variant under 10 CFR 72.48, the NRC issued Severity Level IV violations (Reference 2) that indicated that the design variant should have resulted in a request for amendment to the HI-STORM 100 Certificate of Compliance (CoC) 72-1014. Specifically, the NRC determined that the non-mechanistic tip-over analysis performed for the CBS design included changes to elements of a previously approved method of evaluation (MOE) as well as the use of a new or different MOE thus requiring prior NRC approval. It is unknown when an NRC approved MOE for non-mechanistic tip-over analysis of the MPC-68MCBS can be expected. As such, CEG requests approval of this exemption request by May 9, 2025, to support the next loading campaign to include MPC-68MCBS canisters which is scheduled to begin on June 9, 2025.The attachment to this letter provides the justification and rationale for the exemption request.

There are no regulatory commitments contained in this submittal.

LaSalle County Station 10 CFR Part 72 Exemption Request September 19, 2024 Page 2

If you have any questions or require additional information, please contact Christian Williams at (267) 533-5724.

Respectfully,

Wendi Para Sr. Manager - Licensing Constellation Energy Generation, LLC

Attachment:

Constellation Request for Specific Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 for LaSalle County Station

cc: w/ Attachment Regional Administrator - NRC Region III Resident/Senior Resident Inspector - LaSalle County Station NRC Project Manager - LaSalle County Station Illinois Emergency Management Agency - Division of Nuclear Safety Attachment

CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR LASALLE COUNTY STATION

CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR LASALLE COUNTY STATION

I. Description

The Holtec International Inc., (Holtec) HI-STORM 100 dry cask storage system is designed to hold, and store spent fuel assemblies for independent spent fuel storage installation (ISFSI) deployment. The system is listed in 10 CFR 72.214 as Certificate of Compliance (CoC) Number 72-1014 (Reference 1). This system is scheduled for use by Constellation Energy Generation, LLC (CEG) at LaSalle County Station (LSCS) in accordance with 10 CFR 72.210, General license issued.

Pursuant to 10 CFR 72.7, Specific Exemptions, CEG requests an exemption from the requirements of 10 CFR 72.212(a)(2), 10 CFR 72.212(b)(3), 10 CFR 72.212(b)(5)(i), 10 CFR 72.212(b)(11), and 10 CFR 72.214 for the LSCS Integrated Spent Fuel Storage Installation (ISFSI). Specifically, an exemption is requested for the Holtec 68M Multi-P urpose Canisters with a Continuous Basket Shim (MPC-68MCBS) design basis condition requiring analysis of a postulated non-mechanistic tip-over event using NRC approved methods of evaluation (MOE).

The requested exemption will allow continued storage of loaded storage casks with MPC-68CBS canisters, as listed in Table 1. Additionally, t he requested exemption will allow future loading of MPC-68MCBS canisters, as listed in Table 2.

The exemption is needed because although Holtec performed a non-mechanistic tip-over analysis with favorable results and subsequently implemented the CBS design variant under 10 CFR 72.48, the NRC issued Severity Level IV violations (Reference 2) that indicated that the design variant should have resulted in an amendment to the HI-STORM 100 CoC 72-1014.

Specifically, the NRC determined that the non-mechanistic tip-over analysis performed for the CBS design included changes to elements of a previously approved MOE as well as the use of a new or different MOE thus requiring prior NRC approval. On August 9, 2024, Holtec submitted the application to amend CoC 72-1014 to address the issues identified in Reference 2. This amendment, when adopted at LSCS, will restore the impacted casks to full compliance.

Based on the submit date of August 9, 2024, CEG does not anticipate the amendment will be available for use prior to being needed as LSCS. As such, CEG requests approval of this exemption request by May 9, 2025, to support the next loading campaign to include MPC-68MCBS canisters which is scheduled to begin on June 9, 2025.

The technical justification supporting use of the MPC-68MCBS is provided in the following sections.

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Table 1: List of Affected Canisters Currently Loaded

HI-STORM Serial MPC Serial Number Location on LSCS Date Placed in Number ISFSI Pad Storage 100S(218.0)-1663 811 80 7/13/23 100S(218.0)-1664 812 21 7/20/23 100S(218.0)-1665 813 36 7/31/23 100S(218.0)-1666 814 66 8/3/23

Table 2: List of Affected Canisters Scheduled for Loading

HI-STORM Serial MPC Serial Number Targeted Location Date Targeted to be Number on LSCS ISFSI Pad Placed in Storage 100S(218.0)-1667 815 82 7/18/25 100S(218.0)-1668 816 23 7/25/25 100S(218.0)-1669 817 38 8/1/25 100S(218.0)-1670 818 68 8/8/25

II. Background

LSCS currently utilizes the HI-STORM 100 System under CoC No. 72-1014, Amendment No. 8, Revision 1 for dry storage of spent nuclear fuel in specific Multi-P urpose Canisters (MPC) (i.e.,

MPC-68M canisters). All design features and contents must fully meet the HI-STORM 100 CoC, operations must occur within the Limiting Conditions for Operations (LCOs), and the site must demonstrate that it meets all site-specific parameters.

Holtec International is the designer and manufacturer of the HI-STORM 100 system. Holtec developed a variant of the design for the MPC-68M known as MPC-68MCBS. The MPC-68MCBS basket, like the previously certified MPC-68M, is made of Metamic-HT, and has the same geometric dimensions and assembly configuration. Improvements implemented through the new variant pertain to the external shims which are between the basket periphery and the MPC shell, and the elimination of the difficult to manufacture friction-stir-weld (FS W) seams joining the raw edges of the basket panels.

The CBS variant calls for longer panels of Metamic-HT. The projections of the Metamic-HT panels provide an effective means to secure the shims to the basket using a set of stainless-steel fasteners. These fasteners do no t carry any primary loads, except for the dead weight of the shims when the MPC is oriented vertically, which generates minimal stress in the fasteners.

The fasteners are made of Alloy X stainless material, which is a pre-approved material for the MPCs in the HI-STORM 100 system. Fixing the shim to the basket has the added benefit of improving the heat transfer path from the stored fuel to the external surface of the MPC.

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Holtec performed a non-mechanistic tip-over analysis with favorable results and subsequently implemented the CBS design variants under 10 CFR 72.48. However, the NRC issued Severity Level IV violations (Reference 2) that indicated that these design variants should have resulted in an amendment to the HI-STORM 100 CoC, 72-1014.

A multi-disciplinary team of thermal, criticality, shielding, and structural NRC reviewers assessed a potential structural failure of the fuel basket during accident conditions for the HI-STORM 100 and HI-STORM Flood/Wind (FW) dry cask storage systems to determine the safety significance of these violations. The conclusions were documented and made public in NRC Memorandum, Safety Determination of a Potential Structural Failure of the Fuel Basket During Accident Conditions for the HI-STORM 100 and HI-STORM Flood/Wind Dry Cask Storage Systems, (Reference 3).

III. Basis for Approval of Exemption Request

In accordance with 10 CFR 72.7, the NRC may, upon application by an interested person or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.

a) Authorized by Law

This exemption would allow LSCS to load the MPC-68MCBS design canisters during future loading campaigns. The NRC issued 10 CFR 72.7 under the authority granted to it under Section 133 of the Nuclear Waste Policy Act of 1982, as amended, 42 U.S.C. § 10153. Section 72.7 allows the NRC to grant exemptions from the requirements of 10 CFR Part 72. Granting the proposed exemption will not endanger life or property, or the common defense and security, and is otherwise in the public interest. Therefore, the exemption is authorized by law.

b) Will not Endanger Life or Property or the Common Defense and Security

The NRC has performed a safety assessment (Reference 3) to evaluate the loading and storage of the MPC-68MCBS variant without an NRC approved tip-over analysis. This evaluation (detailed below) assumed basket failure due to the non-mechanistic tip-over event but [] concluded that the consequences of a basket failure have a very low safety significance provided the confinement boundary is maintained and the fuel is kept in a dry storage condition. As these conditions are demonstrated to be met during a tip-over event, the [NRC] staff determined that there was no need to take an immediate action with respect to loaded HI-STORM FW and HI-STORM 100 dry cask storage systems with the continuous basket shim (CBS) fuel basket designs. Based on the NRC safety assessment detailed below and summarized here, the proposed exemption does not endanger life or property or the common defense and security.

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c) Otherwise in the Public Interest

It is in the publics interest to grant an exemption, since dry storage places the fuel in an inherently safe, passive system, and the exemption would permit the continued storage of already loaded canisters before full compliance. This exemption would also allow LSCS to execute scheduled loading campaigns to move spent fuel from the LSCS Fuel Pools to dry storage before full compliance. This exemption would maintain the ability to offload fuel from the reactor, thus allowing continued safe reactor operation.

The following LSCS-specific information is being provided to demonstrate that this exemption is otherwise in the public interest.

Maintain Full Core Discharge Capabilities:

The most significant impact of not being able to use CBS type canisters in upcoming campaigns relates to the ability to effectively manage the margin to full core discharge capability (FCDC) in the LAS Spent Fuel Pool (SFP).

The following margin discussion is based on anticipated loading schedules, which are not controlled documents, and should be considered estimates or targets.

Currently, LAS has a FCDC margin of 315 open cells in the SFP. Following the 2025 refueling outage (L2R20), the FCDC margin will be reduced to 47 open cells due to a planned discharge of 268 fuel assemblies to the SFP. Loading eight (8) MPC-89 canisters (4 CBS-type and 4 FSW-type) in the 2025 Spent Fuel Loading Campaign (SFLC) will increase this margin to 591 open cells. The 2026 refueling outage (L1R21) will decrease the FCDC margin to 323 open cells due to a planned discharge of 268 fuel assemblies to the SFP. Since LAS doesnt have a SFLC scheduled in 2026, removing the four (4) CBS-type canisters from the 2025 SFLC will result in a low margin to FCDC following the 2026 refueling outage (L1R21) and a complete a loss of FCDC following the 2027 refueling outage (L2R21). The FCDC margin will remain negative until the 2027 SFLC.

Having low margins to FCDC makes it difficult to stage the complete reload batch of fuel in the SFP in preparation for outages. This presents a potential reactivity management risk to fuel handling operations during pre-and post-outage.

Decay Heat Removal Requirements:

Each spent fuel bundle contributes to the decay heat removal demand on the spent fuel pool cooling systems. The estimated decay heat from the spent fuel that is scheduled to be moved to dry storage is 1 to 2% per cask. Additionally, removing spent fuel bundles from the fuel pool allows for dispersion of the remaining heat load.

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Accident Consequences and Probability:

Design Bases Accidents associated with the fuel pool include a loss of fuel pool cooling event and a fuel handling accident (FHA). The consequence of a loss of fuel pool cooling is made worse due to the 1 to 2% additional decay heat load contributing to increasing fuel pool temperatures as well as the additional spent fuel experiencing the loss of cooling.

The consequence of an FHA is not impacted however the likelihood of an FHA is increased based on additional fuel moves required to manage fuel pool loading with extra bundles in to pool.

Margin to Capacity:

Once spent fuel pool capacity is reached, the ability to refuel the operating reactor is limited thus taking away a highly reliable clean energy source.

Logistical Considerations and Cascading Impact:

Cask Loading campaigns are budgeted, planned, and scheduled years in advance of the actual performance. LSCS has already purchased and received four (4) MPC-68MCBS canisters with the additional four (4) non-CBS canisters in the manufacturing pipeline.

The current order for non-CBS canisters was initiated based on the expectation that the CoC amendment for the CBS design would not be approved prior to this campaign.

CEG Campaigns are scheduled based on the availability of the specialized work force and equipment that is shared throughout the CEG fleet. These specialty resources support multiple competing priorities including refueling outages, loading campaigns, fuel pool cleanouts, fuel inspections, fuel handling equipment upgrades and maintenance, fuel sipping, new fuel receipt, and crane maintenance and upgrades. Each of these activities limit the available windows to complete cask loading campaigns and delays in any one of these activities has an obvious cascading impact on all other scheduled specialized activities.

==

Conclusion:==

Maintaining adequate FCDC margin ensures operational flexibility necessary for sustained safe and efficient operation of the operating nuclear facility.

Additionally, based on the logistic and financial impact on CEG as discussed above when compared to the minimal safety benefit discussed in the NRC safety memo,

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delaying the use of the MPC-68MCBS canisters does not provide a measurable public benefit.

In contrast, approval of the referenced exemption request supports the continued safe, efficient, and cost-effective operation of LSCS and is therefore in the publics interest.

IV. Technical Justification

The MPC-68MCBS basket assembly features the same fuel storage cavity configuration as the certified standard MPC-68M configuration. The manner in which the inter-panel connectivity is established and by which the aluminum shims are held in place outside the basket is improved. This improvement is made such that, the loose aluminum shims around the basket periphery used in the original MPC-68M design are replaced with integrated aluminum shims that are mechanically fastened (bolted) to basket panel extensions that protrude into the annular region between the basket and the enclosure vessel. The addition of these bolted shims eliminates the need for the FSW located in the external periphery of the Metamic-HT fuel basket. All other fuel basket design characteristics are unchanged by using the CBS variant.

Regardless of their design, the primary design functions of the basket shims are to facilitate heat transfer away from the fuel basket and spent fuel assemblies and to provide lateral support of the fuel basket during the non-mechanistic tip over accident. The primary design functions of the Metamic-HT fuel basket itself, regardless of shim configuration, are to provide structural support of the fuel assemblies and perform the criticality control design function for the system. The MPC enclosure vessel provides structural support of the fuel basket, assisting in the heat transfer process, and acts as the confinement boundary for the system.

Thermal

The NRC staff used the structural assessment discussed below to confirm there was no loss of confinement integrity and considered the thermal impacts of a postulated non-mechanistic tip-over accident. The staff considered fuel debris that might cause hot spots near the bottom of the MPC (on its side from a postulated tip-over). The staff noted that there might be some local increase in temperatures, but no temperatures that would challenge the MPC confinement based on its stainless-steel material. The thermal review concluded, [...] the containment will remain intact and therefore the non-mechanistic tip-over accident condition does not result in significant safety consequences for the HI-STORM FW and HI-STORM 100 storage systems.

Structural and Confinement

The hypothetical tip-over accident is the most significant challenge of the structural performance of the basket. The primary safety function is to prevent a criticality event, and as stated below, the criticality assessment determined no safety concerns under a hypothetical tip-over including basket failure.

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The staff assessment (Reference 3) concluded that the MPC, which is the confinement boundary, maintains its structural integrity during a tip-over event and therefore no water can enter the interior of the MPC during accident conditions. The staff also acknowledged that, consistent with the FSAR, there is no requirement to demonstrate structural integrity of the cladding. Retrievability requirements continue to be met since, as stated above, the MPC maintains its integrity.

The staff also considered natural phenomena hazards (NPH) and concluded, [] the structural failure of the fuel baskets during these NPH accident conditions is unlikely.

However, even if a basket failure occurs, the criticality evaluation below demonstrates that the fuel will be maintained subcritical. Therefore, the staff concludes that the NPH accident conditions do not result in significant safety consequences for the HI-STORM FW and HI-STORM 100 storage systems with the CBS fuel basket designs, (Reference 3).

Finally, the structural assessment considered the handling operations for the dry cask storage systems. The system is either handled with single failure proof devices where a drop is considered non-credible or held to a lift height which has been demonstrated to be acceptable via a drop analysis. The drop analysis shows that there are no significant loads on the basket that would challenge the structural integrity. The NRC concluded that [...] a similar conclusion to that for the non-mechanistic tip-over can be made for dry cask handling accident conditions.

The MPC confinement boundary maintains its structural integrity and no water can enter the interior of the MPC. (Reference 3)

The following is taken from the LSCS 72.212 Evaluation Report, Revision 9

Section 1.1 Conditions of the CoC

[]

Section 1.1.4 Condition 5 - Heavy Loads Requirements

Condition 5 of the CoC requires cask lifts to be performed in accordance with existing plant heavy load requirements and procedures.

[]

Section 9.1.2.2.3 of the LSCS UFSAR specifies administrative controls applicable to handling heavy loads including spent fuel casks in the Reactor Building. These administrative controls are implemented through procedures LFS-400- 1, Reactor Building Overhead Crane 0HC02G Critical L-Path Surveillance Test Prior to Cask Handling Operations in the Restricted Cask Mode, LFP-800- 13, Operation of 125-10 Ton Reactor Building Overhead Crane, and MA-AA-716- 022, Control of Heavy Loads Program (References 17, 18, and 32).

The Reactor Building Overhead Crane (RBOC) is single-failure proof. The RBOC Trolley has been upgraded to a single-failure proof crane in accordance with NUREG-0554, NUREG-0612 (Reference 37), and ASME NOG-1 Edition 2004 (Reference 64) to allow the crane to be used

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for spent fuel cask handling. The main hoist is classified as a Type-1 main hook per NOG-1 and is single-failure proof up to 125 Tons. The auxiliary hook is rated at 10 Ton and is also single-failure proof. Testing of the crane has been performed in accordance with Engineering Change 369785. Since the crane is single failure proof and the HI-TRAC/MPC lifting yokes and devices comply with ANSI N14.6 (Reference 38), the requirements for the components used for cask lifting and transfer inside the Fuel Handling Building in accordance with NUREG-0612 (Reference 37) have been met.

Condition 5 of the CoC also requires plant-specific regulatory reviews (under 10 CFR 50.59 or 10 CFR 72.48) for the heavy load lifts to be performed to show operational compliance with existing plant-specific heavy loads requirements. 10 CFR 50.59 evaluations demonstrating that use of the HI-STORM 100 Cask System is in compliance with existing LSCS heavy loads requirements have been performed. These evaluations are listed in Section 5 of this report.

Condition 5 of the CoC also requires that lifting operations outside of structures governed by 10 CFR 50 must be in accordance with Section 5.5 of Appendix A and/or Sections 3.4.6 and Section 3.5 of Appendix B to the CoC, as applicable. LSCS is in compliance with Sections 5.5 of CoC Appendix A and 3.4.6.a of Appendix B as described in Subparagraph 1.2.3.2 of this report.

LSCS does not have a Cask Transfer Facility (CTF); therefore Section 3.5 of Appendix B is not applicable.

Based on the above, lifts of the HI-STORM 100 Cask System at LSCS will be made in accordance with existing heavy load requirements and procedures. LSCS is therefore in compliance with Condition 5 of the CoC.

[]

Section 1.2 CoC 1014 Appendix A - Technical Specifications Compliance

[]

Section 1.2.3.2 Section 5.5 - Cask Transport Evaluation Program

Section 5.5 establishes requirements for the site transportation of a loaded HI-STORM overpack or HI-TRAC transfer cask.

A loaded HI-TRAC is never transported outside structures governed by 10 CFR 50; therefore section 5.5 does not apply to LSCS for this activity.

Transportation of a loaded HI-STORM 100 overpack into and out of the Reactor Building is provided by a low-profile transporter (LPT) with Hilman rollers that provides support from underneath; therefore Section 5.5 does not apply to LSCS for this activity.

Transportation of loaded HI-STORM 100 overpack between the Reactor Building and the ISFSI is accomplished by a VCT. The VCT is a lifting device designed in accordance with ANSI N14.6 (Reference 38) and employing redundant drop protection features (References 25 and 36).

Accordingly, LSCS is in compliance with subsection 5.5.a.3 Appendix A.

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The applicable procedures governing the activities discussed in Section 1.2.3.2 are listed below.

• LFP 0800- 13 Operation of 125-10 Ton Reactor Building Overhead Crane.

• LFP 0 800- 64 Transporter Operations

• LFP 0800- 65 Spent Fuel Cask Site Transportation

• LFP 0800- 69 HI-TRAC Movement within the Reactor Building

• LFP 0800- 70 HI-TRAC Loading Operations

• LFP 0800- 82 MPC Unloading Operations

Shielding and Criticality

In Reference 3, the NRC staff assessed the potential for a criticality incident under a complete failure of the basket, which could result in basket material and fuel debris at the bottom of the MPC. The staff relied on documented studies related to the enrichment of uranium needed to achieve criticality in an unmoderated, unreflected environment. The allowable contents have enrichment limits well below that in the studies and would also still have the neutron absorbing material present. Therefore, the staff concluded [] there is no criticality safety concern for the CBS basket variants for both the HI-STORM 100 and FW casks under the assumption of fuel basket failure.

As documented in Reference 3, the NRC staff reviewed the shielding impact and concluded,

[] as the damage is localized and the vast majority of the shielding material remains intact, the effect on the dose at the site boundary is negligible. Therefore, the site boundary doses for the loaded HI-STORM FW overpack for accident conditions are equivalent to the normal condition doses, which meet the Title 10 of the Code of Federal Regulations (10 CFR) Section 72.106 radiation dose limits. This statement is applicable to the HI-STORM 100 overpack as the bases for the statement also applies to the HI-STORM 100 system.

Materials

There is no change in the materials used in the CBS variant of the basket compared to the original design of the MPC and basket. Therefore, there is no new material related safety concern.

Safety Conclusion

The above analysis demonstrates that structural failure of the CBS basket resulting from a non-mechanistic tip-over event does not endanger life or property or the common defense and security.

As such the safety significance of using an approved non-mechanistic tip-over analysis completed without using NRC approved methods of evaluation, is bounded by the analysis summarized and discussed in this request which assumed structural basket failure during the postulated event.

V. Environmental Consideration

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The proposed exemption does not meet the eligibility criterion for categorical exclusion for performing an environmental assessment as set forth in 10 CFR 51.22(c)(25) because the exemption does not satisfy the requirement of 10 CFR 51.22(c)(25)(vi). Specifically the request does not involve exemption from any of the following requirements: (A)

Recordkeeping requirements; (B) Reporting requirements; (C) Inspection or surveillance requirements; (D) Equipment servicing or maintenance scheduling requirements; (E)

Education, training, experience, qualification, requalification or other employment suitability requirements; (F) Safeguard plans, and materials control and accounting inventory scheduling requirements; (G) Scheduling requirements; (H) Surety, insurance or indemnity requirements; or (I) Other requirements of an administrative, managerial, or organizational nature.

LSCS has evaluated the environmental impacts of the proposed exemption request and has determined that neither the proposed action nor the alternative to the proposed action will have an adverse impact on the environment. Therefore, neither the proposed action nor the alternative requires any Federal permits, licenses, approvals, or other entitlements.

a) Environmental Impacts of the Proposed Action

The LSCS ISFSI is a radiologically controlled area on the plant site. The area considered for potential environmental impact because of this exemption request is the area in and surrounding the ISFSI.

The interaction of a loaded HI-STORM 100 system with the environment is through thermal, shielding, and confinement design functions for the cask system.

In Reference 3 the NRC documented the following conclusion:

A non-mechanistic tip-over accident condition is considered a hypothetical accident scenario and may affect the HI-STORM FW overpack by resulting in limited and localized damage to the outer shell and radial concrete shield. As the damage is localized and the vast majority of the shielding material remains intact, the effect on the dose at the site boundary is negligible. Therefore, the site boundary doses for the loaded HI-STORM FW overpack for accident conditions are equivalent to the normal condition doses, which meet the Title 10 of the Code of Federal Regulations (10 CFR)

Section 72.106 radiation dose limits.

The LSCS Radiation Shielding Analysis demonstrating compliance with 10 CFR 72.104 is documented in Section 3.1 of the LSCS 72.212 Evaluation Report Revision 17 (Reference 4). The results of the shielding analysis are provided in Section 3.1.3.

Regarding compliance with 10 CFR 72.106, Section 11.2.3.3 of the HI-STORM 100 Final Safety Analysis Report, Revision 11.1 (Reference 5) demonstrates that there are no accidents which would significantly affect shielding effectiveness of the HI-STORM

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100 system and that the requirements of 10 CFR 72.106 are easily met by the HI-STORM 100 system for the postulated tip-over event.

The distance from the ISFSI fence to the Controlled Area Boundary is 768.4 meters as documented in Section 3.1.3 of the LSCS 72.212 evaluation report (Reference 4),

which exceeds the 100-meter minimum distance specified in 10 CFR 72.106.

Based on the above and the NRCs conclusion that damage is localized and the vast majority of the shielding material remains intact, compliance with 10 CFR 72.104 and 10 CFR 72.106 is not impacted by a non-mechanistic tip-over event resulting in basket failure. Therefore, compliance is not impacted by approving the subject exemption request.

There are no gaseous, liquid, or solid effluents (radiological or non-radiological),

radiological exposures (worker or member of the public) or land disturbances associated with the proposed exemption. Therefore, approval of the requested exemption has no impact on the environment.

b) Adverse Environmental Effects Which Cannot be Avoided Should the Exemption be Approved

Since there are no environmental impacts associated with approval of this exemption, there are no adverse environmental effects which cannot be avoided should the exemption request be approved.

c) Alternative to the Proposed Action

In addition to the proposed exemption request, alternative action has been considered.

CEG has considered requesting that Holtec provide eight (8) fully compliant canisters for the June 2025 campaign however, CEG expects that Holtec will not be able to meet this request based on current manufacturing schedules. Additionally, CEG considered delaying the start of the 2025 campaign until Amendment 19 of the HI-STORM 100 CoC is approved. This delay would result in the LSCS ISFSI campaign needing to pull specialty resources which would otherwise be required to support other CEG Refueling and ISFSI campaigns.

d) Environmental Effects of the Alternatives to the Proposed Action

There are no environmental impacts associated with the alternative to the proposed action.

e) Environmental Conclusion

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As a result of the environmental assessment, the future use of MPC-68MCBS at LSCS is in the public interest in that it ensures timely transition of spent fuel to the preferred dry storage facilities and maximizes operational flexibility.

VI. Conclusion

As the safety assessment and environmental review above demonstrate, the HI-STORM 100 system with the MPC-68MCBS canister is capable of performing required safety functions and is capable of mitigating the effects of design basis accidents. Therefore, use of an approved non-mechanistic tip-over analysis completed without using NRC approved methods of evaluation does not present a threat to public and environmental safety.

CEG has reviewed the requirements in 10 CFR 72 and determined that an exemption to certain requirements in 72.212 and 72.214 are necessary. This exemption request would allow the continued storage and future loading of the Holtec HI-STORM 100 MPC-68MCBS systems currently in non-compliance for the term specified in the CoC. The exemption provided herein meets the requirements of 10 CFR 72.7.

References

1 HI-STORM 100 Certificate of Compliance 72-1014 Amendment No. 8, Revision 1, effective 2/16/2016 (ML16041A233) 2 EA-23- 044: Holtec International, INC. - Notice of Violation; The U.S. Nuclear Regulatory Commission Inspection Report No. 07201014/2022-201, ML24016A190 3 NRC Memorandum, Safety Determination of a Potential Structural Failure of the Fuel Basket During Accident Conditions for the HI-STORM 100 and HI-STORM Flood/Wind Dry Cask Storage Systems, dated January 31, 2024, ML24018A085 4 LaSalle County Station 10 CFR 72.212 Evaluation Report, Revision 9 5 HI-STORM 100 Final Safety Analysis Report, Revision 11.1

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