ML24082A008

Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214

ML24082A008
Person / Time
Site:	Calvert Cliffs, 07200078, 07201032
Issue date:	03/22/2024
From:	David Gudger Constellation Energy Generation
To:	Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation, Document Control Desk
References
Download: ML24082A008 (1)

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Subject:

Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 For Calvert Cliffs Nuclear Power Plant - Holtec MPC-37CBS

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 Calvert Cliffs Nuclear Power Plants Independent Spent Fuel Storage Installation ( ISFSI).

Specifically, an exemption is requested for the Ho lt e c 37 M ulti-P urpose Canisters (MPC) with a Continuous Basket Sh im (MPC-37CBS) design basis condition requiring analysi s of a postulated non-mechanistic tip-over event.

The requested exemption will allow continued storage of loaded storage casks with MPC -

37CBS canisters, as listed in Table 1. Additionally, the exemption will allow future loading of MPC-37CBS canisters, as listed in Table 2.

The exemption is needed becaus e 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 FW CoC 72-1032. 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. It is unknown when an NRC approved MOE for non-mechanistic tip-over analysis of the MPC -37CBS can be expected. As such, CEG requests approval of this exemption req uest by July 12, 2024, to support the next loading campaign to include MPC-37CBS canisters which is scheduled to begin on July 15, 2024.The attachment to this letter provides the justification and rationale for the exemption request.

There are no regulatory comm itm ents contained in this subm ittal.

If you have any questions or require additional information, please contact Christian Williams at (267) 533-5 724.

Calvert Cliffs Nuclear Power Plant 10 CFR Part 72 Exem ption Request March 22, 2024 Page 2 of 3

Respectfully,

David T. Gudger Sr. Manager, Licensing Constellation Energy Generation, LLC

Attachm ent: Constellation Request for Specific Exemption From Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 fo r Calvert Cliffs Nuclear Power Plant

cc: w/ Attachment Regional Administrator - NRC Region I Resident/Senior Resident Inspector - Calvert Cliffs Nuclear Power Plant NRC Project Manager - Calvert Cliffs Nuclear Power Plant

Attachm ent

CONSTELLATION REQUEST FOR S PECIFIC EXE MPTION FRO M CERTAIN RE QUIRE MENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QU AD CITIES NUCLEAR POWER STATION

CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR CALVERT CLIFFS NUCLEAR POWER PLANT

I. Description

The Holtec International Inc., (Holtec) Storage Module Flood and Wind ( HI-STORM FW) 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) Num ber 72-1032 (Reference 1). This system is used by Constellation Energy Generation, LLC (CEG) at Calvert Cliffs Nuclear Power Plant (CCNPP) in accordance with 10 CFR 72. 210, General License Issued.

Pursuant to 10 CFR 72.7, Specific Exemptions, CEG requests an exem ption 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 fo r th e CCNPP Independent Spent Fuel Storage Installation (ISFSI). Specifically, an exemption is requested for the Holtec 37 Multi-P urpose Canisters with a Continuous Basket Shim (MPC-37CBS) 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-37CBS canisters, as listed in Table 1. Additionally, the exem ption will allow future loading o f MPC-37CBS canisters, as listed in Table 2.

The exemption is needed because although Ho lt ec performed a non-mechanistic t ip-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 an amendment to the HI-STORM FW Co C 72-1032. Specifically, the NRC determ ined that th e 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.

It is unknown when an NRC approved MOE for non-mechanistic tip-over analysis of the MPC-37CBS can be expected. As such, CEG requests approval of this exemption request by July 12, 20 2 4, to support the next loading campaign to include MPC-37CBS canisters which is scheduled to begin on Ju ly 15, 2024.

The technical justification supporting continued use of the MPC-37CBS 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 ISFSI Date Placed in Number Pad Storage 154 169 12 16 November 2021 153 168 20 9 Novem ber 2021 152 170 28 19 November 2021 355 241 18 2 Septem ber 2022 356 242 26 12 September 2022 357 243 34 20 September 2022 358 244 16 22 August 2023 384 298 24 25 August 2023 385 299 32 1 Septem ber 2023

Table 2: List of Affected Canisters Scheduled for Loading

HI-STORM Serial MPC Serial Number Targeted Location Date Targeted to be Number on ISFSI Pad Placed in Storage 386 300 14 18 July 2024 425 301 22 25 July 2024 426 302 30 1 August 2024 427 303 03 8 August 2024 428 304 04 15 August 2024 429 305 05 22 August 2024

II. Background

CCNPP currently utilizes the HI-STORM FW System under CoC No. 72-1032, Am endment No.

1, Revision No. 1 (Reference 1) for dry storage of spent nuclear fuel in specific Multi-P urpose Canisters (MPC) (i.e., MPC-37 canisters). All design features and contents must fully meet the HI-STORM FW CoC, operations must m eet the specified Limiting Conditions for Operations (LCOs), and the site must demonstrat e that it m eets all site-specific parameters.

Holtec International is the designer and manufacturer of the HI -S TO RM FW system. Holtec developed a variant of the design for the MPC -37 known as MPC -37CBS. The MPC-37CBS basket, like the previously certified MPC -37, is made of Metamic -HT, and has the sam e 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 (F SW) seams joining the raw edges of the basket panels.

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The CBS variant calls for longer panels of Metamic -HT. The projections of the Metamic -HT panels provide an effect ive m eans 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 FW 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.

Holtec performed a non-mechanistic t ip-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 a proposed amendment to the HI -STORM FW CoC, 72-1032 to request NRC approval of the use of a new or different MOE and changes to elements of a previously approved MOE.

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 FW dry cask storage systems to determine the safety significance of these violations. The conclusions were documented and made public in NRC Mem orandum,

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 a n 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 CCNPP to continue to store previously loaded and load additional canisters of the MPC -37CBS design. 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 n ot 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 perform ed a safety assessment (Reference 3) to evaluate the loading and storage of the MPC -37CBS variant without an NRC approved tip-over analysis. This evaluation (detailed below) assumed basket failure due to the non-mechanistic tip-over event and [] concluded that the consequences of a basket failure have a very low

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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 summar ized here, the proposed exemption does not endanger life or property or the common defense and security.

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 th e continued storage of already loaded canisters before full compliance. This exemption would also allow upcoming loading campaigns to proceed on time to move fuel into the dry storage condition and maintain the ability to offload fuel from the reactor, thus allowing continued safe reactor operation.

The following CCNPP-specific inform ation is being provided to further 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 CCNP Un it 1 and CCNP Unit 2 Spent Fuel Pools (SFP).

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

Currently, CCNP has a FCDC margin of 48 open cells in the SFP. Loading six (6) MPC -

37 canisters in the 2024 Spent Fuel Loading Campaign (SFLC) will increase this margin to 270 open cells. The 2025 refueling outage (CC2R26) will decrease the FCDC margin to 174 open cells due to a planned discharge of 96 fuel assemblies. If CAL removes all six (6) MPC-37 canisters from the 2024 SFLC scope, the site will lose FCDC (margin -

48) following CC2R26. Since C CNP doesnt hav e a SFLC scheduled in 2025, the FCDC margin will remain at -48 until the 2026 refueling outage (CC1R28). The 2026 refueling outage (CC1R28) will decrease the FCDC margin to -144 due to a planned discharge of 96 fuel assemblies. CCNP would need to load t en (10) MPC-37 canisters in the 2026 SFLC to restore FCDC margin.

Loss of FCDC for two years of CAL Unit 1 and Unit 2 operation presents unnecessary risks/challenges to SFP inventory and SFP operations.

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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. Also, divers routinely perform underwater maintenance on SFP fuel transfer system in preparation for refueling outages. To maintain dose to divers as low as reasonably achievable, SFP fuel racks adjacent to the fuel transfer system are emptied of spent fuel. Having low or no margin to FCDC, could result in higher dose to divers.

Decay Heat Removal Requirements:

Each spent fuel bundle contributes to the decay heat removal demand on the SFP 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 SFP allows for dispersion of the remaining heat load.

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 S FP capacity is reached, the ability to refuel to 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. Cam paigns are scheduled based on the availability of the specialized work force and equipment that is shared throughout the CEG fleet. T hese specialty resources support multiple competing priorities including refueling outages,

loading cam paigns, fuel pool cleanouts, fuel inspections, fuel handling equipment upgrades and maintenance, fuel sipping, new fuel receipt, and crane maintenance an d 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.

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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 assessment (Reference 3), delaying the use of the MPC -37CBS 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 CCNPP and is therefore in the publics interest.

IV. Technical Justification

The MP C-37CBS basket assembly features the same fuel storage cavity configuration as the certified standard MPC -37 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 i n the original MPC -37 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 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 conclude d, [...] the containment will remain intact and therefore the non-mechanistic tip-over accident condition

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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.

The NRC safety assessment (Reference 3) concluded that the MPC, which is the confinem ent 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 acknowledge d th at, consistent with the HI-STORM FW Final Safety Analysis Report (Reference 5), 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 int egrity.

The NRC 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 conclu des 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 operatio ns 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 CCNPP 72.212 Evaluation Report, Revision 3 (Reference 2)

1.1 Conditions of the CoC

[]

1.1.4 Condition 4 - Heavy Loads Requirements

[]

Statement of Compliance:

Lifts of the HI-TRAC VW transfer cask and the MPC are performed using single-failure-proof handling systems as defined by Section 5.1.6 of NUREG -0612. The spent fuel cask handling

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crane (a single-failure-proof crane and integral structure governed by 10 CF R Part 50 regulations) along with the applicable special lifting devices and interfacing lift points are the three major components that establish a single-failure-proof handling system. Additional information on the various configurations of the single-failure-proof handling systems utilized at Calvert Cliffs are provided in Table 1-3 and discussed in Section 1.2.5.2 of this report. Calvert Cliffs UFSAR Section 5.7 (Ref. 5) describes the spent fuel cask handling crane as being designated single-failure-p roof using criteria from NUREG -0554, Single-Failure-Proof Cranes For Nuclear Power Plants (Ref. 9). All lifts with this crane are performed in accordance with MA-AA-716-021, Rigging and Lifting Program (Ref. 14), and MA-AA-716-022, Control of Heavy Load s Program (Ref. 15). Lifts involving an empty MPC in the Auxiliary Building are also performed in accordance with the plants heavy load control program (Ref. 15). Finally, lifts of the HI-STORM FW overpack are not performed using the spent fuel cask handling crane or any other structure governed by 10 CFR Part 50 requirements, therefore lifts of the HI -S TO RM FW overpack are discussed in Section 1.2.5.2 of this report.

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1.2 CoC No. 1032-1 R1 Appendix A - Technical Specifications Compliance

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1.2.5.2 Section 5. 2 - Transport Evaluation Program

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Statement of Compliance:

Table 1-3 of this report provides the various transport configurations of a loaded MPC, HI -TRAC VW, and HI-STORM FW at Calvert Cliffs and determines the applicability of the governing CoC requirements and ensures that all requirements are captured.

The loaded MPC and HI -TRAC VW are only handled using single-failure-proof handling systems as defined by Section 5.1.6 of NUREG- 0612. The spent fuel cask handling crane (a single-failure-proof crane and integral structure governed by 10 CFR Part 50 regulations) along with the applicable special lifting devices and interfacing lift points are the three major components that establ ish a single-failure-proof handling system, (Ref 5). Additional information on how the various configurations of these single-failure-proof handling systems are utilized at Calvert Cliffs and how they meet the guidance of Section 5.1.6 of NUREG -0612 are pr ovided in Table 1-3. Additionally, the HI -STORM FW overpack is not lifted with equipment that is integral to a structure governed by 10 CFR Part 50. Therefore, Subsection 5.2.a is satisfied.

The loaded HI -TRAC VW is never transported outside of structures that are governed by 10 CFR 50. The transfer of the loaded MPC from the HI -TRAC VW to the HI -STORM FW overpack will occur within the Auxiliary Building. This process, MPC Transfer, is specifically defined by CoC 1032-1R1 to begin when the MPC is lifted off the HI-TRAC VW bottom lid and end when the MPC is supported from beneath by the HI -STORM FW (or the reverse). The MPC is lifted off

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of the HI-TRAC VW bottom lid through the use of the Lift Lock, a special lifting device, designed to the increased stress limits of ANSI N14.6, (Ref. 7). The Lift Lock uses an adapter to interface with the crane hook of the spent fuel cask handling crane.

Transportation of the loaded HI -STORM FW overpack out of the Auxiliary Building to the ISFSI is provided by a self -propelled modular transporter, known as the HI -PORT. The HI-PORT consists of two heavy duty modules connected by a center drop deck, which utilizes a hydraulic system for leveling and is operated with diesel fuel. The HI -PORT supports the HI-S TO RM overpack from underneath. Therefore, per Subsection 5.2.b, Section 5.2 does not apply to Calvert Cliffs when the HI-STORM FW overpack is transported and supported by the HI -PORT.

Once the HI -PORT with the loaded HI-STORM FW is just outside the ISFSI pad on Camp Canoy road, a Vertical Cask Transporter (VCT) is used to lift the loaded HI -STORM FW from the HI-PORT. The VCT then enters the ISFSI and is used to place the HI -STORM FW in its final storage location on the ISFSI pad. The VCT and lifting attachments are devices th at are designed to the requirements of Section 5.2.c of Appendix A of the CoC and Section 1.2.1.5 of the HI-STORM FW FSAR (Ref. 2). The VCT that is utilized at Calvert Cliffs has been designed, fabricated, operated, tested, inspected, and maintained to meet the requirements of Sections 5.2 c.1. through 5.2.c.3., as described in RRTI -2845-003 (Ref. 61).

The applicable CE G procedures governing these activities are listed below.

• OU-CA-630-200, MPC Loading at Calvert

• OU-CA-630-300, MPC Processing at Calvert

• OU-CA-630-400, MPC Transfer at Calvert

• OU-CA-630-500, HI-STORM Movements at Calvert

• OU-CA-630-600, MPC Unloading at Calvert

Shielding and Criticality

In their safety assessm ent (Reference 3), the NRC 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 environm ent. 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 the safety assessment ( Reference 3), th e NRC reviewed the shielding impact and concluded, [] as the dam age 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.

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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 assum ed structural basket failure during the postulated event.

V. Environmental Consideration

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). Speci fically 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.

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

The interaction of a loaded HI -STORM FW 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:

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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, th e 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.

CCNPP Effluents and Direct Radiation associated with the ISFSI are discussed in Section 3 of the CCNPP 72.212 Evaluation Report Revision 3 (Reference 4).

3.1 ISFSI Radiation Shielding Analysis

Holtec Report HI -2188620, Calvert Cliffs Site Boundary Dose Rates Calculations for HI-STORM FW System (Ref. 4), provides estimated dose values for the HI -STORM FW casks in storage at the Calvert Cliffs ISFSI.

[]

3.2 Confinement Analysis

[]

The Holtec HI-STORM FW FSAR, (Ref. 2), as well as the NRC SER for the CoC, (Ref.

1), concludes that there are no credible design basis events that would result in a radiological release from the MPCs. Therefore, no confinement analysis has been performed for C alvert Cliffs and a non-mechanistic effluent release contribution has not been added to the Calvert Cliffs ISFSI dose rate calculations.

[]

3.3.1 10 CFR 72.104(a) - Limits

[]

Statement of Compliance:

Calculations were performed to determine the expected maximum annual dose rates as a result of storing the additional HI -STORM FW casks on the Calvert Cliffs ISFSI in Holtec Report HI -2188620, Calvert Cliffs Site Boundary Dose Rates Calculations for HI-STORM FW System (Ref. 4).

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The calculated annu al dose rates for a full time occupancy factor (8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />) to any real individual beyond the controlled area [1000 meters from the pad with the cask array depicted in (Ref. 7)] is summarized in Table 3-1. It is important to note that the controlled area boundary is located 1189 meters from the pad, and therefore the considered distance of 1000 meters in HI -2188620 (Ref. 4) is conservative.

The contribution of the maximum annual dose rate from both the NUHOMS System and from reactor site operations is provided in the Annual Radioactive Effluent Release Report for the Calvert Cliffs Nuclear Power Plant and Independent Spent Fuel Storage Installation (Ref. 12). Table 3-1 demonstrates that the dose rate at the controlled area boundary is below the limits mentioned in 10 CFR 72.104(a).

3.3.2 10 CFR 72.104(c) - Dose Control

[]

Statement of Compliance:

As discussed in Chapter 7 of the HI -STORM FW FSAR, the MPCs containing spent fuel are seal -welded and tested to meet leak -tight criteria before being placed into service at the ISFSI. The MPCs are designed to maintain confinement integrity under all normal, off normal and accident events. Therefore, no effluent limits are established. Further, as discussed in Section 3.1 of this report, direct radiation doses are ensured by analysis to be less than the limits of 10 CFR 72.104(a). Loading of casks only with authorized contents and deployment of the casks in accordance with the CoC and FSAR ensure that regulatory dose limits are not approached.

Environmental monitoring and periodic reporting confirm that doses due to ISFSI operation are below those in 10 CFR 72.104(a).

Regarding compliance with 10 CFR 72.106, Section 11. 4.3 of the HI-STORM FW Final Safety Analysis Report, Revision 3 (Reference 5) demonstrates that there are no accidents which would significantly affect shielding effectiveness of the HI -STORM FW system and that the requirements of 10 CFR 72.106 are easily met by the HI -STORM FW system for the postulated tip-over event.

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 sol id effluents (radiological or non-radiological),

radiological exposures (worker or member of the public) or land disturbances

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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.

Specifically, the existing MPC-37CBS canister would need to be unloaded and re-loaded into the older design MPC -37 canisters. Future loading campaigns would also need to be delayed until older design canisters can be fabricated and delivered to site.

In addition, the reflooding of the MPCs, removal of fuel assemblies, and replacement into a different MPC would result in additional doses and handling operations with no added safety benefit, since it has been demonstrated that the MPC maintains all its safety functions.

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

As a result of the environmental assessment, the continued storage and future use of MPC-37CBS at CCNPP is in the public interest in that it avoids unnecessary additional operations and incurred dose that would result from the alternative to the proposed action.

VI. Conclusion

As the safety assessment and environmental review above demonstrate, the HI-STORM FW system with th e MP C-37CBS canister is capable of performing required safety functions and is capable of mitigating the effects of design basis accidents (DBAs ). 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

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allow the continued storage and future loading of the Holtec HI-STORM FW MP C-37CBS 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 FW Certificate of Compliance 72-1032 Am endm ent No. 1, Revision No.

1, effective June 2, 2015 (ML15152A399) 2 EA-23-044: Holtec International, INC. - Notice of Violation; The U.S. Nuclear Regulatory Commission Inspection Report No. 072010 14/2022-201 (ML24016A190 ), dated January 30, 2024 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 (ML24018A085), dated January 31, 2024 4 Calvert Cliffs Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 3, effective December 2023 5 HI-STORM FW Final Safety Analysis Report, R evision 4, dated June 24, 2015

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