L-24-076, Independent Spent Fuel Storage Installation - Supplemental Information for Request for Specific Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214
ML24082A132 | |
Person / Time | |
---|---|
Site: | Perry, 07200069 |
Issue date: | 03/22/2024 |
From: | Penfield R Vistra Operations Company |
To: | Office of Nuclear Reactor Regulation, Document Control Desk |
References | |
L-24-076 | |
Download: ML24082A132 (1) | |
Text
Rod L. Penfield Site Vice President 10 Center Road Perry, Ohio 44081 440- 280-5382
L 076 10 CFR 72.7 March 22, 2024
ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555- 0001
SUBJECT:
Perry Nuclear Power Plant, Unit No. 1 Docket No. 50- 440, License No. NPF -58 Perry Nuclear Power Plant, Unit No. 1 Independent Spent Fuel Storage Installation Docket No. 72-69 Supplemental Information for Request for Specific Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214
On February 27, 2024, Energy Harbor Nuclear Corp requested U.S. Nuclear Regulatory Commission (NRC) approval for Specific Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 (ML24058A180).
Effective March 1, 2024, the facility operating license for PNPP was transferred from Energy Harbor Nuclear Generation LLC (owner) and Energy Harbor Nuclear Corp.
(operator) to Energy Harbor Nuclear Generation LLC (owner) and Vistra Operations Company LLC (ML24057A092). Upon completion of this license transfer, Vistra Operations Company LLC (VistraOps) assumed the responsibility for all licensing actions under NRC review at the time of the transfer and requested that the NRC continue its review of these actions (ML24054A498).
On March 6, 2024, the NRC held a public meeting to discuss the path forward for general licensees affected by the Holtec continuous basket shims basket design change. In response to the NRC meeting, VistraOps is providing supplemental information to address information discussed in the meeting. Specifically, further
Perry Nuclear Power Plant, Unit No. 1 L-24-0 76 Page 2
description of the administrative controls in place to minimize the risk of a cask tip-over and confirmation that the dose limits in 10 CFR 72.104 and 10 CFR 72.106 are met.
Additionally, clarifications are added to the Environmental Consideration section.
The attachment to this letter provides the justification and rationale for the exemption request. To simplify the review, the attachment is a total rewrite of the previously submitted request.
VistraOps requests approval of this exemption by July 1, 2024, to support the upcoming summer 2024 dry cask loading campaign.
There are no regulatory commitments contained in this submittal. If there are any questions, or if additional information is required, please contact Jack Hicks, Sr Manager, Licensing, at (254) 897-6725 or Jack. Hicks@luminant.com.
Sincerely,
Rod L. Penfield
Attachment:
Request for Specific Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214
cc: NRC Region III Administrator NRC Resident Inspector NRR Project Manager Utility Radiological Safety Board Attachment L-24-0 76
Request for Specific Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 Page 1 of 12
TABLE OF CONTENTS
I. DESCRIPTION
II. BACKGROUND
III. BASIS FOR APPROVAL OF EXEMPTION REQUEST
IV. TECHNICAL JUSTIFICATION
V. ENVIRONMENTAL CONSIDERATION
VI. CONCLUSION
VII. REFERENCES
Attachment L-24-0 76 Page 2 of 12
I. Description
The Holtec International Inc., (Holtec) 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 Number 72-1032. This system is used by Vistra Operations Company LLC (VistraOps) at Perry Nuclear Power Plant (PNPP) in accordance with 10 CFR 72.210, General license issued.
Pursuant to 10 CFR 72.7, Specific Exemptions, VistraOps requests an exemption from certain 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 PNPP.
Specifically, an exemption is requested to allow use of the Holtec 89 multi -purpose canisters (MPCs) with a Continuous Basket Shim (MPC -89CBS) design variant. If approved, the requested exemption will allow loading of MPC -89CBS canisters, as listed in Table 1.
The exemption is needed because although Holtec originally performed a tip-over analysis with favorable results and subsequently implemented the CBS design vari ants under 10 CFR 72.48, 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 FW CoC No. 72-1032. Specifically, the 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 new or different MOEs thus requiring prior NRC approval via an amendment, which is not expected to be approved prior to PNPPs upcoming loading campaign.
VistraOps requests approval of this exemption request by July 1, 2024, to support the loading of the next MPC -89CBS canister scheduled for August 2024.
The technical justification supporting continued use of the MPC -89CBS is provided in the following sections.
Table 1: List of Affected Canisters Scheduled for Loading
HI-STORM Serial MPC Serial Targeted Location Date Scheduled to Number Number on ISFSI Pad be Placed in Storage HI-STORM 0283 MPC 0371 Pad 3, location 2 8/16/24 HI-STORM 0284 MPC 0372 Pad 3, location 3 8/23/24
Attachment L-24-0 76 Page 3 of 12
II. Background
PNPP currently utilizes the HI-STORM FW System under Certificate of Compliance No. 72-1032, Amendment No. 5 (CoC) (Reference 1) and the corresponding Holtec HI-STORM FW Final Safety Analysis Report, Rev. 9 (FSAR) (Reference 5), for dry storage of spent nuclear fuel in specific MPCs (that is, MPC -89CBS canisters). All design features and contents must fully meet the CoC requirements, including required MPC or spent fuel contents and technical specification loading requirements within the limiting conditions for operations (LCOs), and the site must demonstrate that they meet all site-specific parameters per 10 CFR 72.212.
Holtec International is the designer and manufacturer of the HI -STORM FW system.
Holtec developed a variant of the design for the MPC -89 known as MPC-89CBS. The MPC-89CBS basket, like the previously certified MPC -89, is made of Metamic -HT, and has the same geom etric 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 (FSW) seams joining the raw edges of the basket panels.
The CBS variant calls for longer panels of Metamic -HT. The projections of the Metamic panels provide an effective means to secure the shims to the basket using a set of stainless -steel fasteners. These fasteners do not 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 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 FW CoC number 1032.
A multi-disciplinary NRC team of thermal, criticality, shielding, and structural staff 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, (staff assessment) (Reference 3).
Attachment L-24-0 76 Page 4 of 12
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 is 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 VistraOps to load additional canisters of the MPC-89CBS 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 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 staff assessment to evaluate the loading and storage of the MPC-89CBS variant without an approved tip-over analysis. This evaluation (detailed in section IV) assumed basket failure due to the 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 staff determined that there was no need to take an immediate action with respect to loaded HI -STORM 100 and HI -STORM FW dry cask storage systems with the continuous basket shim (CBS) fuel basket designs.
Based on the staff assessment detailed below and summarized 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. This exemption would allow the upcoming loading campaign 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 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 offload capability (FCO), that is the number of available cells in excess of FCO, in the PNPP Spent Fuel Pool (SFP).
Attachment L-24-0 76 Page 5 of 12
The following margin discussion is based on anticipated loading schedules, which are not controlled documents and should be considered estimates or targets.
Currently, PNPP has an FCO margin of 21 open cells in the SFP. Loading 6 standard MPC-89 and 2 MPC -89CBS in the 2024 Spent Fuel Loading Campaign (SFLC) will increase this margin to 733 open cells. After 2024, the next SFLC is planned for 2028 when 7 MPC-89CBS canisters are planned to be loaded. If PNPP removes the 2 MPC-89CBS canisters from the 2024 SFLC scope, the margin to FCO decreases to 555 open cells. The upcoming 2025 and 2027 refueling outages (PY1R20 and PY1R21) will ultimately decrease the FCO margin to a deficit of 25 open cells due to the planned discharges of 288 and 292 fuel bundles, respectively. That is, PNPP will lose FCO in 2027 and will not regain FCO, and margin to FCO, until the 2028 SFLC. Having no FCO for over a year of PNPP operation is an unnecessary risk to the operation of the plant, SFP inventory, and SFP operations. In order to regain FCO prior to the 2028 SFLC, non-fuel components stored in the pool would need to be relocated. This effort involves additional resources, dose, and risk to perform.
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. Removing spent fuel bundles from the fuel pool allows for dispersion of the remaining heat load and increases the time to reach SFP temperatures that would require the use of supplemental cooling.
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. Campaigns are scheduled based on the availability of the specialized work force and resources that are shared throughout the VistraOps fleet.
These specialty resources support multiple competing priorities including refueling outages, loading campaigns, fuel rearrangement in the pools, fuel inspections, fuel handling equipment upgrades and maintenance, fuel sipping, new fuel receipt, and crane maintenance and upgrades. Each of these ac tivities 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 FCO 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 VistraOps as discussed above when compared to the minimal safety benefit discussed in the NRC safety Attachment L-24-0 76 Page 6 of 12
memo, delaying the use of the MPC -89CBS 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 PNPP.
IV. Technical Justification
The MPC-89CBS basket assembly features the same fuel storage cavity configuration as the certified standard MPC-89 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 t he basket periphery used in the original MPC -89 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 Metami c-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.
The results of the staff assessment are discussed below for the critical parameters or basic nuclear safety criteria as identified within the 10 CFR 72.48 proces s of the Holtec HI-STORM 100 and HI-STORM 100 Flood/Wind (FW) dry cask storage system with the CBS design variant.
Thermal The 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 100 and HI -STORM FW storage systems.
Attachment L-24-0 76 Page 7 of 12
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 event with the assumption of basket failure.
The staff assessment 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 100 and HI -STORM FW storage systems with the CBS fuel basket designs.
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 that has been demonstrated to be acceptable. Additional discussion on the lifting and handling operations for PNPP is provided below. 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. Should the fuel basket fail to maintain its structural integrity during stack-up the fuel will be maintained in a subcritical condition, (Reference 3).
The lifting and handling operations and requirements for the dry cask system is documented in the PNPP 10 CFR 72.212 Evaluation Report (Reference 4).
Appendix 1, Table 1, Condition 4 evaluates the heavy loads requirements concluding that PNPP is in compliance with Condition 4 of CoC. All lifts of the cask will be conducted in accordance with PNPP's existing heavy load program and administrative controls. PNPP's Fuel handling Building (FHB) crane has been upgraded t o single failure proof. Also, 10 CFR 50.59 evaluations have been performed demonstrating that heavy load lifts and the use of the HI -STORM FW system are in compliance with existing PNPP heavy load requirements.
Appendix 1, Table 2, Section 5.2 evaluates cask transport. This section establishes the requirements for the site transportation of a loaded HI -STORM FW Overpack or HI-TRAC transfer cask. A loaded HI -TRAC is never transported Attachment L-24-0 76 Page 8 of 12
outside structures governed by 10 CFR 50 and the FHB crane used to lift the HI-TRAC is integral to a structure governed by 10 CFR 50. Transportation of a loaded HI-STORM FW Overpack into and out of the FHB is provided by a Zero Profile Transporter (ZPT) with Hilman rollers that provides support from underneath. Transportation of a loaded HI -STORM FW Overpack between the FHB and the ISFSI is accomplished by the Holtec HI -TRAN, which meets the CoC requirements.
Appendix 1, Table 3, Section 3.4 evaluates site-specific parameters and analyses. The site-specific parameters and analyses relative to cask handling and transport that require verification, include:
Overpack Transporter Seismic Stability The HI-STORM FW Version F will be transported outside of the FHB on the ZPT. A comparative analysis of this configuration verses the loaded HI-STORM 100S Version B on the ZPT was performed and concluded that the HI-STORM FW Version F is bounded by the analysis of the HI-STORM 100S Version B. The seismic analysis of the HI -STORM 100S Version B on ZPT determines that the ZPT and HI -STORM demonstrate a large factor of safety against overturning under the bounding design basis earthquake. The HI -STORM FW Version F will also be transported to the ISFSI pad using the HI-TRAN Vertical Cask Transporter. The seismic stability of this configuration was evaluated and determined that the HI-TRAN, while carrying the loaded HI -STORM overpack, remains kinematically stable and does not overturn during a seismic event.
HI-STORM/HI-TRAC Stackup Seismic Stability The evaluation of the HI-STORM FW Version F and the HI -TRAC VW transfer cask stackup configuration during a design basis seismic event shows the HI-STORM FW system is bounded by the HI -STORM 100 system, based on stack weight and dynamic parameters of HI -STORM FW system. Therefore, the conclusions made for the HI -STORM 100 system stack, which show the stack remains kinematically stable, apply to the HI-STORM FW system at PNPP.
Also, bounding lift heights for all cask transport operations at PNPP have been determined. The evaluated lift heights are identified as restrictions in the applicable loading procedures for PNPP.
Applicable procedures governing the handling and transport activities are listed below and available for inspection if required.
HPP-3119- 0200, MPC Loading at Perry HPP-3119- 0300, MPC Processing at Perry HPP-3119- 0400, MPC Transfer at Perry HPP-3119- 0500, HI-STORM Movements at Perry HPP-3119- 0600, MPC Unloading at Perry HPP-3119- 0700, Responding to Abnormal Condition at Perry Attachment L-24-0 76 Page 9 of 12
Shielding and Criticality In the staff assessment, the 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 the staff assessment, the NRC staff reviewed the shielding impact and concluded, 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.
PNPP documents compliance with 10 CFR 72.104 in Section 5.3.1.6 of the 72.212 Report stating that the, Radiological Evaluation demonstrates by analysis that there is reasonable assurance the annual dose equivalent to any real individual who is located beyond the ISFSI controlled area boundary during normal operations and anticipated off -normal occurrences is not to exceed 10 CFR 72.104(a) dose limits for the combination of doses from the ISFSI and PNPP.
Regarding compliance with 10 CFR 72.106, Section 12.2 of the FSAR 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.
The minimum distance from the ISFSI to the Site boundary, which also serves as the Owner Controlled Area boundary, is documented in Section 5.3.1.4 of the 72.212 Report. The minimum distance is approximately 428 meters which is greater than 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 Attachment L-24-0 76 Page 10 of 12
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.
Radiation Protection As there is no adverse effect on the shielding or confinement functions, there is no effect on occupational or public exposures as a result of thi s off-normal event.
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 tip-over event does not endanger life or property or the common defense and security. As such the safety significance of not having an approved tip-over analysis, demonstrating the structural integrity of the CBS design during the postulated tip-over event, is bounded by the analysis assuming structural basket failure.
V. Environmental Consideration
The proposed exemption does not meet the eligibility criterion for categorical ex clusion 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 admini strative, managerial, or organizational nature.
PNPP 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 t he 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 PNPP 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.
Attachment L-24-0 76 Page 11 of 12
The interaction of a loaded HI -STORM FW system with the environment is through thermal, shielding, and confinement design functions for the cask system. Based on the safety analysis described above, the following conclusions for the proposed storage of the MPC-89CBS variant have been verified.
- The confinement boundary maintains its structural integrity during accident conditions.
- Fuel cladding temperature limits will not challenge NUREG -2215 limits.
- Existing radiological evaluations and conclusions for accident conditions in Chapter 5 of the FSAR remain valid, since, consistent with the above analysis, these evaluations assume localized damage and the majority of the shielding material remains intact.
Further, there are no discussions or analyses in the 72.212 Report that would contradict or negate the conclusion made in this 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 env ironment.
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, future loading campaigns with MPC -89CBS canisters would need to be delayed until older design canisters can be fabricated and delivered to site.
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 future use of MPC -89CBS canisters at PNPP is in the public interest in that it avoids unnecessary delays and additional activities that would result from the alternative to the proposed action.
Attachment L-24-0 76 Page 12 of 12
VI. Conclusion
As the safety assessment and environmental review above demonstrate, the HI-STORM FW system with the MPC-89CBS canister is capable of performing required safety functions and is capable of mitigating the effects of desi gn 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.
VistraOps has reviewed the requirements in 10 CFR 72 and determined that an exemption to certain requirements in 72.212 and 72.214 are necessary. The exemption provided herein meets the requirements of 10 CFR 72.7. This exemption request would allow future loading of MPC -89CBS canisters, as listed in Tabl e 1.
VII. References
1 HI-STORM FW Certificate of Compliance 72-1032 Amendment No. 5, effective 7/27/2020, ML20163A701.
2 EA-23- 044: Holtec International, INC. - Notice of Violation; The U.S. Nuclear Regulatory Commission Inspection Report No. 07201032/2022-201, 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 Perry Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 2.
5 HI-STORM FW Final Safety Analysis Report, Revision 9.