RS-24-020, Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 for Quad Cities Nuclear Power Station - Holtec MPC-68MCBS
| ML24075A001 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities, Holtec  |
| Issue date: | 03/15/2024 |
| From: | David Gudger Constellation Energy Generation |
| To: | Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| RS-24-020 | |
| Download: ML24075A001 (1) | |
Text
200 Exelon Way Kennett Square, PA 19348 www.constellation.com 10 CFR 72.7 RS-24-020 March 15, 2024 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Quad Cities Nuclear Power Station Unit 1 and 2 Renewed Facility Operating Licensee Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254, 50-265, and 72-53
Subject:
Request for Exemption from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 For Quad Cities Nuclear Power 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 Quad Cities Independent Spent Fuel Storage Installation (ISFSI). Specifically, an exemption is requested for the Holtec 68M 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 loading of MPC-68MCBS canisters during the upcoming campaign scheduled to begin in June of 2024.
The exemption is needed because although Holtec originally performed a non-mechanistic tip-over analysis with favorable results and subsequently implemented the CBS design variants 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 100 Certificate of Compliance (CoC) number 72-1014. Specifically, 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 new or different MOEs thus requiring prior NRC approval. It is unknown when an NRC approved MOE for non-mechanistic tip-over analysis of the MPC-68MCBS would be expected. As such, CEG requests approval of this exemption request by June 21, 2024, to support the next loading campaign to include the MPC-68MCBS canisters which is scheduled to begin on June 24, 2024.
The attachment to this letter provides the justification and rationale for the exemption request.
There are no regulatory commitments contained in this submittal.
Quad Cities Nuclear Power Station 10 CFR Part 72 Exemption Request March 15, 2024 Page 2 of 3 If you have any questions or require additional information, please contact Christian Williams at (267) 533-5724.
Respectfully, David T. Gudger 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 Quad Cities Nuclear Power Station cc:
w/ Attachment Regional Administrator - NRC Region III Resident/Senior Resident Inspector - Quad Cities Nuclear Power Station NRC Project Manager - Quad Cities Nuclear Power Station
Attachment CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QUAD CITIES NUCLEAR POWER STATION
CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QUAD CITIES NUCLEAR POWER STATION Page 1 of 11 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 Quad Cities Nuclear Power Station (QCNPS) 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 QCNPS Integrated Spent Fuel Storage Installation (ISFSI). Specifically, an exemption is requested for the Holtec 68M Multi-Purpose 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 loading of MPC-68MCBS canisters, as listed in Table 1.
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. It is unknown when an NRC approved MOE for non-mechanistic tip-over analysis of the MPC-68MCBS would be expected.
As such, CEG requests approval of this exemption request by June 21, 2024, to support the next loading campaign to include MPC-68MCBS canisters which is scheduled to begin on June 24, 2024.
The technical justification supporting use of the MPC-68MCBS is provided in the following sections.
CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QUAD CITIES NUCLEAR POWER STATION Page 2 of 11 Table 1: List of Affected Canisters Scheduled for Loading HI-STORM Serial Number MPC Serial Number Targeted Location on QCNPS ISFSI Pad 2 Date Targeted to be Placed in Storage 100-1706 829 78 11 July 2024 100-1707 830 77 18 July 2024 100-1708 831 76 25 July 2024 100-1709 832 75 01 August 2024 II.
Background
QCNPS 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-Purpose 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 (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-HT 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 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.
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
CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QUAD CITIES NUCLEAR POWER STATION Page 3 of 11 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 QCNPS 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.
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 QCNPS to execute scheduled loading campaigns to move spent fuel from the QCNPS 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.
CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QUAD CITIES NUCLEAR POWER STATION Page 4 of 11 The following QCNPS-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 QCNPS Unit 1 and Unit 2 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, QCNPS has a FCDC margin of 42 open cells in the SFP. Loading three (3)
MPC-68M and four (4) MPC-68MCBS in the 2024 Spent Fuel Loading Campaign (SFLC) will increase this margin to 550 open cells. The 2025 refueling outage (Q1R28) will decrease the FCDC margin to 298 open cells due to a planned discharge of 220 fuel bundles. If QCNPS removes the four (4) MPC-68MCBS canisters from 2024 SFLC scope, the FCDC margin will drop to only 26 open cells following Q1R28. Since QCNPS doesnt have a SFLC scheduled in 2025, the FCDC margin will remain at 26 until the 2026 RFO. The 2026 refueling outage (Q2R28) will decrease the FCDC margin to -198 (loss of FCDC) due to a planned discharge of 224 fuel bundles.
Having a FCDC margin of only 26 open cells for over a year of QDC operation and the loss of FCDC in 2026 present unnecessary risks/challenges to spent fuel pool (SFP) inventory and SFP operations.
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.
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
CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QUAD CITIES NUCLEAR POWER STATION Page 5 of 11 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 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. 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, 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 QCNPS 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
CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QUAD CITIES NUCLEAR POWER STATION Page 6 of 11 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.
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
CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QUAD CITIES NUCLEAR POWER STATION Page 7 of 11 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 QCNPS 72.212 Evaluation Report, Revision 17 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 9.1.4.2.2, Reactor Building Overhead Crane, of the QCNPS UFSAR specifies administrative controls applicable to handling heavy loads including spent fuel casks in the reactor building. These administrative controls are implemented through procedures QCFHP 0500-16, Reactor Building Crane Surveillance Test Prior to Operations in the Restricted Mode, QCMM 5800-05, Reactor Building Overhead Crane Utilization, and MA-AA-716-022, Control of Heavy Loads Program (References 17, 18, and 19).
Section 9.1.4.3.2, Reactor Building Overhead Crane, of the QCNPS UFSAR describes the single-failure-proof design of the reactor building overhead crane. This single-failure-proof design was reviewed and found acceptable (Reference 34) by the Franklin Research Center (NRC contractor). Further, the NRC states in Reference 35 that the reactor building overhead crane is licensed as single-failure-proof for loads up to 110 tons.
[]
Section 1.2 CoC 1014 Appendix A - Technical Specifications Compliance
[]
Section 1.2.3.2 Section 5.5 - Cask Transport Evaluation Program
CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QUAD CITIES NUCLEAR POWER STATION Page 8 of 11 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 HI-TRAC activities in this 72.212 Evaluation Report.
Transportation of a loaded HI-STORM 100 overpack into and out of the Reactor Building is provided by a low-profile transporter with Hillman rollers that provides support from underneath[.] Therefore, Section 5.5 does not apply to QCNPS for this activity.
Transportation of a loaded HI-STORM 100 overpack between the Reactor Building and the ISFSI is accomplished by a Vertical Cask Transporter (VCT). The VCT is a lifting device designed in accordance with the increased safety factors of ANSI N14.6 and employing redundant drop features (References 29 and 30). Accordingly, QCNPS is in compliance with Section 5.5.a.3 and the overpack may be lifted to any height necessary during transport operations.
The applicable procedures governing the activities discussed in Section 1.2.3.2 are listed below.
QCFHP 0800-64 TRANSPORTER OPERATIONS QCFHP 0800-65 SPENT FUEL CASK SITE TRANSPORTATION QCFHP 0800-68 HI-TRAC PREPARATION QCFHP 0800-69 HI-TRAC MOVEMENT WITHIN THE REACTOR BUILDING QCFHP 0800-70 HI-TRAC LOADING OPERATIONS QCFHP 0800-82 MPC UNLOADING OPERATIONS QCMM 5800-05 REACTOR BUILDING OVERHEAD CRANE UTILIZATION 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.
CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QUAD CITIES NUCLEAR POWER STATION Page 9 of 11 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 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.
QCNPS 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 QCNPS 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:
CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QUAD CITIES NUCLEAR POWER STATION Page 10 of 11 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 QCNPS Radiation Shielding Analysis demonstrating compliance with 10 CFR 72.104 is documented in Section 3.1 of the QCNPS 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 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 conservatively estimated to be 250 meters, based on publicly available maps, 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
CONSTELLATION REQUEST FOR SPECIFIC EXEMPTION FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 and 10 CFR 72.214 FOR QUAD CITIES NUCLEAR POWER STATION Page 11 of 11 In addition to the proposed exemption request, alternative action has been considered.
Specifically, future loading campaigns 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-68MCBS at QCNPS 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 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
Quad Cities Nuclear Power Station 10 CFR 72.212 Evaluation Report, Revision 17 5
HI-STORM 100 Final Safety Analysis Report, Revision 11.1