ML22077A132

Request for Exemption from 10CFR50.54(w)(1) Re Onsite Property Damage Insurance

ML22077A132
Person / Time
Site:	Indian Point
Issue date:	03/18/2022
From:	Fleming J Holtec Decommissioning International
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
HDI-IPEC-22-025
Download: ML22077A132 (23)

Text

I...-H 0 LT E C Krishna P. Singh Technology Campus, 1 Holtec Blvd., Camden, NJ 08104 Telephone (856) 797-0900 DECOMMISSIONING INTERNATIONAL Fax (856) 797-0909

HDI-IPEC-2 2-0 25 10 CFR 50.12 10 CFR 50.54(w)(1)

March 18, 2022

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

Indian Point Energy Center Provisional License No. DPR-5 Renewed Facility License No. DPR-26 and DPR-64 NRC Docket Nos.50-003, 50-247, 50-286, and 72-051

Subject:

Request for Exemption from 10 CFR 50.54(w)(1) Concerning Indian Point Energy Center Onsite Property Damage Insurance

References:

1. Entergy letter to U.S. NRC, "Notification of Unit 1 Transfer of 160 Spent Fuel Assemblies from the Spent Fuel Pool to the Indian Point Independent Spent Fuel Storage Installation," (Accession No. ML083510667), dated December 11, 2008

2. Entergy letter to U.S. NRC, "Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel, Indian Point Nuclear Generating Unit No. 2," (Accession No. ML20133J902), dated May 12, 2020

3. Entergy letter to U.S. NRC, "Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel, Indian Point Nuclear Generating Unit No. 3," (Accession No. ML21131A157), dated May 11, 2021

4. Holtec Decommissioning International, LLC (HDI) letter to U.S. NRC, "Supplement to HDI Request for Exemptions from Certain Emergency Planning Requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E, Indian Point Nuclear Generating Unit Nos. 1, 2, and 3 Including Site-Specific Calculations," (Accession Nos ML22032A017 and ML22032A027),

dated February 1, 2022

In accordance with Title 10 of the Code of Federal Regulations (CFR) 50.12, "Specific exemptions," Holtec Decommissioning International, LLC (HDI), on behalf of Holtec Indian Point 2, LLC (IP1 & IP2) and Holtec Indian Point 3, LLC (IP3), collectively referred to as Indian Point Energy Center (IPEC), requests permanent exemption from 10 CFR 50.54(w)(1) for IP1, IP2, and IP3. The provisions of 10 CFR 50.54(w)(1) require the licensee to obtain insurance coverage from private sources to provide protection covering the licensee's obligation, in the unlikely event of an accident, to stabilize and decontaminate the reactor and the reactor site. Specifically, licensees must obtain insurance having a minimum coverage limit for each reactor station site of either $1.06 billion or whatever amount of insurance is generally available from private sources, whichever is less. This insurance coverage is referred to as "onsite coverage" or "onsite insurance coverage."

I...-H 0 LT E C Krishna P. Singh Technology Campus, 1 Holtec Blvd., Camden, NJ 08104 Telephone (856) 797-0900 DECOMMISSIONING INTERNATIONAL Fax (856) 797-0909

HDI is requesting exemption from 10 CFR 50.54(w)(1) to reduce the IPEC minimum onsite insurance coverage to $50 million. The exemption request is provided in the Enclosure to this letter.

Operation of IP1 was suspended on October 31, 1974, and all fuel was removed from the reactor vessel in 1975. On December 11, 2008, Entergy notified the NRC that all remaining spent fuel assemblies had been removed from the IP1 spent fuel pool (SFP) and placed in the existing Independent Spent Fuel Storage Installation (ISFSI)

(Reference 1).

In References 2 and 3, Entergy certified to the NRC, in accordance with 10 CFR 50.82(a)(1)(i), that power operations ceased at IP2 on April 30, 2020, and at IP3 on April 30, 2021. In addition, Entergy certified in accordance with 10 CFR 50.82(a)(1)(ii), that the fuel was permanently removed from the IP2 reactor vessel and placed in the IP2 SFP on May 12, 2020, and that the fuel was permanently removed from the lP3 reactor vessel and placed in the IP3 SFP on May 11, 2021.

The underlying purpose of the 10 CFR 50.54(w)(1) is to require sufficient property damage insurance to ensure adequate funding of onsite post-accident recovery, stabilization, and decontamination costs following an accident at an operating nuclear power plant. However, the regulation does not take into consideration the reduced potential for, and consequences of, such nuclear incidents at permanently shutdown facilities. The proposed exemption would allow a reduction in the level of onsite insurance coverage to a level that is commensurate with the permanently shutdown and defueled status of all three IPEC reactors, while meeting the underlying purpose of the rule.

Holtec has performed analyses showing that 15 months after IP3s shutdown, the spent fuel stored in each of the unit's SFPs will have decayed sufficiently such that there is a significant reduction in risk from a theoretical SFP draindown event. This reduction in risk supports the basis for the proposed exemption from 10 CFR 50.54(w)(1) provided in the Enclosure to this letter. The analyses establishing the 15-month spent fuel decay time was provided to the NRC in Reference 4.

The 15-month spent fuel decay period for IP3 will expire by August 1, 2022. Therefore,

HDI requests N RC review and approval of the requested exemption b y July 1, 2022,

with an effective date of August 1, 2022 and a 30-day implementation period from the effective date.

This letter contains no new regulatory commitments.

If you have any questions or need further information, please contact Mr. Walter Wittich, IPEC Licensing or myself at (856) 797-0900, ext. 3578.

Sincerely,

Jean A. Fleming HDI Vice President, Regulatory and Environmental Affairs Holtec Decommissioning International, LLC I...-H 0 LT E C Krishna P. Singh Technology Campus, 1 Holtec Blvd., Camden, NJ 08104 Telephone (856) 797-0900 DECOMMISSIONING INTERNATIONAL Fax (856) 797-0909

Enclosure:

Request for Exemption from 10 CFR 50.54(w)(1)

cc: NRC Senior Project Manager, NRC NMSS NRC Region l Regional Administrator NRC Senior Regional Inspector, Indian Point Energy Center New York State Liaison Officer Designee, NYSERDA New York State (NYS) Public Service Commission Enclosure HDI-IPEC-22-025 Page 1 of 20

Enclosure

HDI-22-025

Request for Exemption from 10 CFR 50.54(w)(1)

Enclosure HDI-IPEC-22-025 Page 2 of 20

DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGES

I. SPECIFIC EXEMPTION REQUEST

II.BACKGROUND

III. DETAILED DESCRIPTION

IV.BASIS FOR EXEMPTION REQUEST

V. TECHNICAL EVALUATION

VI.JUSTIFICATION FOR EXEMPTION AND SPECIAL

CIRCUMSTANCES

VII.PRECEDENT

VIII.ENVIRONMENTAL ASSESSMENT

IX.CONCLUSION Enclosure HDI-IPEC-22-025 Page 3 of 20

I. SPECIFIC EXEMPTION REQUEST

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.12, "Specific exemptions," Holtec Decommissioning International, LLC (HDI), on behalf of Holtec Indian Point 2, LLC (IP1 & IP2) and Holtec Indian Point 3, LLC (IP3), collectively referred to as Indian Point Energy Center (IPEC), hereby requests permanent exemption from 10 CFR 50.54(w)(1) for IP1, IP2, and IP3. The provisions of 10 CFR 50.54(w)(1) require the licensee to obtain insurance coverage from private sources to provide protection covering the licensee's obligation, in the unlikely event of an accident, to stabilize and decontaminate the reactor and the reactor site.

Specifically, licensees must obtain insurance having a minimum coverage limit for each reactor station site of either $1.06 billion or whatever amount of insurance is generally available from private sources, whichever is less. This insurance coverage is referred to as "onsite coverage" or "onsite insurance coverage."

HDI is requesting exemption from 10 CFR 50.54(w)(1) to reduce the IPEC minimum onsite insurance coverage to $50 million.

II. BACKGROUND

IPEC is located on the east bank of the Hudson River at Indian Point, in the Village of Buchanan, in upper Westchester County, New York. The site is operated by Holtec Decommissioning International, LLC (HDI) and contains facilities located on approximately 239 acres, bounded on the north, south, and east by privately owned land and on the west by the Hudson River. IP2 and IP3 are located north and south, respectively, of IP1, which is in safe storage (SAFSTOR) until subsequent decommissioning. The site is located about 24 miles north of the New York City boundary line. The nearest urban area within 6 miles of the site is the City of Peekskill, New York, which is located approximately 2.5 miles northeast of the IPEC site.

IP1 was permanently shutdown on October 31, 1974, and all spent fuel was removed from the IP1 reactor vessel in 1975. On December 11, 2008, Entergy notified the NRC that all remaining spent fuel assemblies had been removed from the IP1 spent fuel pool (SFP) and placed in the existing Independent Spent Fuel Storage Installation (ISFSI) (Reference 1). The IP1 Provisional Operating License prohibits taking the reactor to criticality or operation of the facility at any power level, and the IP1 Technical Specifications do not allow fuel to be loaded into the reactor core or moved into the reactor containment building without prior review and authorization by the U.S. Nuclear Regulatory Commission (NRC). The IP1 Technical Specifications also preclude fuel from being stored in the IP1 fuel storage area. Based on its current configuration and licensing basis, with no spent fuel stored in the IP1 SFP, there are no postulated Design Basis Accidents (DBAs) that remain applicable to IP1. The IP1 SFP is no longer in use because all spent fuel and other material has been removed, and the IP1 SFP has been drained.

By letters dated May 12, 2020 and May 11, 2021 (References 2 and 3), Entergy certified to the NRC, in accordance with 10 CFR 50.82(a)(1)(i), that power operations ceased at IP2 on April 30, 2020, and at IP3 on April 30, 2021. In addition, Entergy certified in accordance with 10 CFR 50.82(a)(1)(ii), that the fuel was permanently removed from the IP2 reactor vessel and placed in the IP2 SFP on May 12, 2020, and that the fuel was permanently removed from the lP3 reactor vessel and placed in the IP3 SFP on May 11, 2021.

After a reactor is in a permanently defueled condition, the operational focus is with the spent fuel and the spent fuel pool (SFP) cooling systems. In this condition, the spectrum of credible Enclosure HDI-IPEC-22-025 Page 4 of 20 accidents is much smaller than for an operational plant. Further, the IP2 and IP3 certifications of permanent shutdown and defuel, in accordance with 10 CFR 50.82(a)(1)(i) and (ii), the 10 CFR Part 50 licenses for IP2 and IP3 no longer authorize operation of the respective unit's reactor or emplacement or retention of fuel in the unit's reactor vessel. As such, the majority of the design basis accident (DBA) scenarios previously postulated in the safety analyses for the plant are no longer possible and were removed under the provisions of 10 CFR 50.59.

III. DETAILED DESCRIPTION

In accordance with 10 CFR 50.12, "Specific exemptions," HDI requests a permanent exemption from 10 CFR 50.54(w)(1) for IPEC. The 10 CFR 50.54(w)(1) rule requires individual power reactor licensees to obtain insurance coverage from private sources to provide protection covering the licensee's obligation, in the unlikely event of an accident, to stabilize and decontaminate the reactor and the reactor site. Specifically, licensees must obtain insurance having a minimum coverage limit for each reactor station site of either $1.06 billion or whatever amount of insurance is generally available from private sources, whichever is less. This insurance coverage is referred to as "onsite coverage" or "onsite insurance coverage."

HDI is requesting an exemption from 10 CFR 50.54(w)(1) to reduce the minimum coverage limit of 10 CFR 50.54(w)(1) to $50 million for IPEC.

10 CFR 50.54(w)(1) reads as follows:

"(w) Each power reactor licensee under this part for a production or utilization facility of the type described in §§ 50.21(b) or 50.22 shall take reasonable steps to obtain insurance available at reasonable costs and on reasonable terms from private sources or to demonstrate to the satisfaction of the NRC that it possesses an equivalent amount of protection covering the licensee's obligation, in the event of an accident at the licensee's reactor, to stabilize and decontaminate the reactor and the reactor station site at which the reactor experiencing the accident is located, provided that:

(1) The insurance required by paragraph (w) of this section must have a minimum coverage limit for each reactor station site of either $1.06 billion or whatever amount of insurance is generally available from private sources, whichever is less. The required insurance must clearly state that, as and to the extent provided in paragraph (w)(4) of this section, any proceeds must be payable first for stabilization of the reactor and next for decontamination of the reactor and the reactor station site. If a licensee's coverage falls below the required minimum, the licensee shall wi thin 60 days take all reasonable steps to restore its coverage to the required minimum. The required insurance may, at the option of the licensee, be included within policies that also provide coverage for other risks, including, but not limited to, the risk of direct physical damage."

Exemption from 10 C)R 50.54(w)(1) is requested in order to allow reduced insurance coverage commensurate with the significantly reduced risks associated with the permanently shutdown and defueled condition of the three unit I3EC site. +oltec has performed analyses showing that 15 months after I33¶s shutdown, the spent fuel stored in the S)3s will have decayed sufficiently such that there is a significant reduction in risk from postulated events. This reduction in risk supports the basis for this proposed exemption from 10 C)R 50.54(w)(1). The analyses establishing the 15 month spent fuel decay time was provided to the NRC in Reference 4.

Enclosure HDI-IPEC-22-025 Page 5 of 20 IV. BASIS FOR EXEMPTION REQUEST

The underlying purpose of 10 CFR 50.54(w)(1) is to require sufficient property damage insurance to ensure adequate funding of onsite post-accident recovery, stabilization, and decontamination costs following an accident at an operating nuclear power plant. The requirements of 10 CFR 50.54(w)(1) were developed taking into consideration the risks associated with an operating nuclear power reactor, including the potential consequences of a release of radioactive material from the reactor. The onsite insurance coverage must be either $1.06 billion or whatever amount of insurance is generally available from private sources (whichever is less).

This regulation does not take into consideration the reduced potential for and consequences of such nuclear incidents at permanently shutdown facilities. IPEC is a three-unit multiple reactor site, with the three onsite reactors (IP1, IP2 and IP3) permanently shutdown and defueled. The proposed exemption would allow a reduction in the level of onsite insurance coverage for the three-unit IPEC multiple reactor site to a level that is commensurate with the permanently defueled status of all three reactors and the underlying purpose of the 10 CFR 50.54(w)(1) rule.

Although the likelihood of an accident at an operating reactor is small, the consequences can be large, in part due to the high temperatures and pressures of the reactor coolant system as well as the inventory of radionuclides. For permanently shutdown and defueled reactors, such as IP1, IP2, and IP3, nuclear accidents involving the reactors and their associated systems, structures, and components (SSCs) are no longer possible. Furthermore, reductions in the probability and consequences of non-operating reactor nuclear incidents are substantially reduced because: 1) the decay heat from the spent fuel decreases over time, which reduces the amount of cooling required to prevent the spent fuel from heating up to a temperature that could compromise the ability of the fuel cladding to retain fission products; and 2) the relatively short-lived radionuclides contained in the spent fuel, particularly volatile components like iodine and noble gases, decay away, thus reducing the inventory of radioactive materials available for release.

The potential for, and consequences of, nuclear accidents decline substantially after the IPEC reactors are permanently defueled, however, they are not completely eliminated. There are potential onsite and offsite radiological consequences that could be associated with the onsite storage of the spent fuel in the SFP. In addition, a site with multiple permanently shutdown and defueled reactors, such as IPEC, may contain an inventory of radioactive liquids, activated reactor components, and contaminated materials. For purposes of modifying the amount of onsite insurance coverage maintained for a site with multiple permanently shutdown and defueled reactors, the potential radiological consequences of non-operating reactor nuclear incidents are appropriate to consider, despite their very low probability of occurrence.

The NRC has generically evaluated the legal, technical, and policy issues regarding the financial protection requirements for large nuclear power plants that have been permanently shutdown and recommended changes to the power reactor financial protection regulations that would allow licensees to lower onsite insurance levels to $50 million. The results of the NRC evaluations were summarized in SECY-96-256 (Reference 5) and the NRC recommended course of action was approved by the Commission in a Staff Requirements Memorandum (SRM) (Reference 6).

These documents established the basis for the NRC exercising its discretionary authority to specify an appropriate level of onsite insurance coverage for permanently shutdown nuclear power reactors.

In SECY-00-145 (Reference 7) and SECY-0 1-0100 (Reference 8), the NRC discussed additional information concerning SFP zirconium fire risks at decommissioning reactors and associated Enclosure HDI-IPEC-22-025 Page 6 of 20 implications for onsite property damage insurance. Analyzing when spent fuel stored in the SFP is capable of adequate air-cooling is one measure that demonstrates when the probability of a zirconium fire would be exceedingly low.

As discussed in the NRC response to a comment submitted by the Nuclear Energy Institute (NEI) in SECY-00-145 (see "NRC Staff Responses to NEI White Paper Comments on Improving Decommissioning Regulations," page 5, response to Comment 2):

"Since the zirconium fire scenario would be possible for up to several years following shutdown, and since the consequences of such fire are severe in terms of property damage and land contamination, the staff position is that full onsite liability coverage must be retained for five years or until analysis has indicated that a zirconium fire is no longer possible."

In addition, as discussed in the NRC response to another NEI comment in SECY 145 (see "NRC Staff Responses to NEI White Paper Comments on Improving Decommissioning Regulations," page 5, response to Comment 3):

" As discussed above, the staff believes that full insurance coverage must be maintained for 5 years or until a licensee can show by analysis that its spent fuel pool is no longer vulnerable to such [a zirconium] fire."

V. TECHNICAL EVALUATION

IP1 was permanently shutdown on October 31, 1974, and all spent fuel was removed from the IP1 reactor vessel in 1975. On December 11, 2008, Entergy notified the NRC that all remaining spent fuel assemblies had been removed from the IP1 spent fuel pool (SFP) and placed in the existing Independent Spent Fuel Storage Installation (ISFSI) (Reference 1). The IP1 Provisional Operating License prohibits taking the reactor to criticality or operation of the facility at any power level, and the IP1 Technical Specifications do not allow fuel to be loaded into the reactor core or moved into the reactor containment building without prior review and authorization by the U.S.

Nuclear Regulatory Commission (NRC). The IP1 Technical Specifications also preclude fuel from being stored in the IP1 fuel storage area. Based on its current configuration and licensing basis, with no spent fuel stored in the IP1 SFP, there are no postulated Design Basis Accidents (DBAs) that remain applicable to IP1. The IP1 SFP is no longer in use because all spent fuel and other material has been removed, and the IP1 SFP has been drained.

By letters dated May 12, 2020 and May 11, 2021 (References 2 and 3), Entergy certified to the NRC, in accordance with 10 CFR 50.82(a)(1)(i), that power operations ceased at IP2 on April 30, 2020, and at IP3 on April 30, 2021. In addition, Entergy certified in accordance with 10 CFR 50.82(a)(1)(ii), that the fuel was permanently removed from the lIP2 reactor vessel and placed in the IP2 SFP on May 12, 2020, and that the fuel was permanently removed from the lP3 reactor vessel and placed in the IP3 SFP on May 11, 2021.

With the reactor in a permanently defueled condition, the operational focus is with the spent fuel and the spent fuel pool (SFP) cooling systems. In this condition, the spectrum of credible accidents is much smaller than for an operational plant. Further, with the IP2 and IP3 certifications of permanent shutdown and defuel, in accordance with 10 CFR 50.82(a)(1)(i) and (ii), the 10 CFR Part 50 licenses for IP2 and IP3 no longer authorize operation of the respective unit's reactor or emplacement or retention of fuel in the unit's reactor vessel. As such, the majority of the design basis accident (DBA) scenarios previously postulated in the safety analyses for the plant are no longer possible and were removed under the provisions Enclosure HDI-IPEC-22-025 Page 7 of 20 of 10 CFR 50.59. For IP1, due to its current configuration and licensing basis with no spent fuel stored in the IP1 SFP, there are no postulated DBAs that remain applicable to IP1.

Accident Analysis Overview

With the termination of reactor operations and permanent removal of fuel from IP1, IP2 and IP3 reactor vessels, the postulated accidents involving failure or malfunction of the reactor and supporting SSCs are no longer applicable to any of the three IPEC plants.

The HDI submittal, "Revision to HDI Request for Exemptions from Certain Emergency Planning Requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E, Indian Point Nuclear Generating Unit Nos. 1, 2, and 3," (Reference 9) provides information on the disposition of accidents and other incidents of concern. Furthermore, as discussed in Reference 9, based on its current configuration and licensing basis, with no spent fuel stored in the IP1 SFP, there are no postulated DBAs that remain applicable to IP1. The IP1 SFP is no longer in use because all spent fuel has been transferred to the ISFSI and other material removed, and the IP1 SFP has been drained.

Accordingly, the analyses discussed within this section only address the risks associated with the storage of spent fuel in the IP2 and IP3 SFPs. Specific analyses are summarized in the following sections.

A. Consequences of Design Basis Events

The NRC approved the IP2 Permanently Defueled Technical Specifications (PDTS) on April 28, 2020, with the issuance of IP2 License Amendment No. 294 (Reference 10). The license amendment included the statement that the applicable DBAs for IP2 in the permanently defueled condition are: (1) an Fuel Handling Accident (FHA) in the Fuel Storage Building (FSB), (2) an accidental release of waste gas, and (3) an accidental release-recycle of waste liquid.

The NRC approved the IP3 Permanently Defueled Technical Specifications on April 22, 2021, with the issuance of IP3 License Amendment No. 270 (Reference 11) reflecting the permanently shutdown and defueled condition. The IP3 amendment includes the statement that the applicable DBAs for IP3 in the permanently defueled condition are: (1) the FHA in the FHB, (2) an accidental release of waste gas, and (3) an accidental release-recycle of waste liquid.

The limiting DBA for IP2 and IP3 in the permanently defueled condition is the FHA in the FSB. An FHA may occur in the FSB during movement of a fuel assembly. The fuel assembly is moved under water and the accident is assumed to occur when the fuel assembly is damaged. The IP2 and IP3 post-permanent shutdown FHA (Reference 12) was evaluated utilizing the Alternate Source Term (AST) methodology described in Regulatory Guide 1.183 (Reference 13). This analysis did not credit the function of FSB filtration, high-rad alarm, dispersion from the FSB ventilation system, Control Room isolation, or emergency filtration. The analysis credits the decontamination of the 23 feet of water over the fuel assemblies in the SFP with an overall effective decontamination factor of 200, consistent with Regulatory Guide 1.183 (Reference 13).

The analysis indicates that after a decay time of at least 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> (30 days) following permanent cessation of power operations of each unit, the FHA results in an EAB TEDE dose of 0.47 rem (Reference 12), which is below the EPAs early phase PAG criteria of 1 rem TEDE for recommended evacuation.

Enclosure HDI-IPEC-22-025 Page 8 of 20 B. Consequences of Beyond Design Basis Events

Spent Fuel Assembly Heat Up During a Theoretical Drain Down Event

The analyses, provided in Enclosure 1 to Reference 4, compare the heat load limits for the hottest fuel assembly and for a 2X2 group of assemblies stored in each SFP (IP2 and IP3) to a criterion proposed in Commission Paper SECY-99-168, Improving Decommissioning Regulations for Nuclear Power Plants," (Reference 14) that is applicable to offsite emergency response for nuclear power reactors in the decommissioning process. This criterion considers the time for the hottest assembly to heat up from 30°C to 900°C adiabatically. A heatup time of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> from the time the spent fuel is uncovered, was determined to be sufficient to take mitigating actions.

The bounding analyses for the IP2 and IP3 SFPs (Reference 4) for beyond design basis events demonstrate that 15 months after shutdown of IP3 a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> is available before the fuel cladding temperature of the hottest fuel assembly in either SFP reaches 900°C with a complete loss of SFP water inventory. As stated in NUREG -1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," (February 2001) (Reference 15) 900°C is an acceptable temperature to use for assessing the onset of fission product release under transient conditions (to establish the critical decay time for determining availability of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to evacuate) if fuel and cladding oxidation occurs in air.

Because of the length of time it would take for the fuel to heatup, there is ample time to respond to any draindown event that might cause such an occurrence by restoring cooling or makeup or providing spray to the IP2 or IP3 SFPs. As a result, the likelihood that such a scenario would progress to a zirconium fire is deemed not credible.

C. Consequences of Other Analyzed Events

Spent Fuel Pool Draindown Event

NRC NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," (Reference 16) Supplement 1, Section 4.3.9, identifies that a SFP draindown event is a beyond design basis event. The bounding analyses for the IP2 and IP3 SFP draindown event (Reference 4) demonstrate that a significant release of radioactive material from the SFP is not possible within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> from the time the spent fuel is uncovered. However, the potential exists for radiation exposure if shielding of the fuel in the IP2 or IP3 SFP is lost.

HDI analyzed the bounding radiological consequences of a postulated complete loss of SFP water from either the IP2 and IP3 SFPs. The analysis considered the distances from both SFPs to both Control Rooms and the EAB and a combination of IP3 fuel in the IP2 SFP, to bound both units (Reference 17).

The SFP water and the concrete SFP structures serve as radiation shielding. Therefore, a loss of water shielding above the fuel could increase the offsite radiation levels because of the gamma rays streaming up out of the SFP and being scattered back to a receptor at the site boundary. The analysis determined that the limiting dose rate in the IP2 and IP3 Control Rooms at one year after shutdown are less than 0.0259 mrem/hr (Reference 26) and the dose rate to a receptor at the EAB is less than 11.55 mrem/hr (Reference 17), which is less than the EPA PAGs.

Enclosure HDI-IPEC-22-025 Page 9 of 20 Consequences of a Beyond-Design Basis Earthquake

NUREG-1738 (Reference 15) identifies beyond design basis seismic events as the dominant contributor to events that could result in a loss of SFP coolant that uncovers fuel for plants in the Central and Eastern United States. Additionally, NUREG-1738 identifies a zirconium fire resulting from substantial loss-of-water inventory from the SFP, as the only postulated scenario at a decommissioning plant that could result in significant offsite radiological release. The scenarios that lead to this condition have very low frequencies of occurrence (i.e., on the order of one to tens of times in a million years) and are considered beyond design basis events because the SFP and attached systems are designed to prevent a substantial loss of coolant inventory under accident conditions. However, the consequences of such accidents could potentially lead to an offsite radiological dose in excess of the EPA PAGs (Reference 18) at the EAB.

The risk associated with zirconium cladding fire events decreases as the spent fuel ages. As the spent fuel ages, the decay time increases, the decay heat decreases, and the short-lived radionuclides decay away. As the decay time increases, the overall risk of zirconium cladding fire continues to decrease due to two factors: (1) the amount of time available for preventative actions increases, which reduces the probability that the actions would not be successful; and (2)the increased likelihood that the fuel is able to be cooled by air, which decreases the reliance on actions to prevent a zirconium fire. The results of the research conducted for NUREG-1738 and NUREG-2161, "Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor," (September 2014) (Reference 19) suggests that, while other radiological consequences can be extensive, a postulated accident scenario leading to a SFP zirconium fire, where the fuel has had significant decay time, will have little potential to cause offsite early fatalities due to dose, regardless of the type of offsite response.

The purpose of NUREG-2161 (Reference 19) was to determine if accelerated transfer of older, colder spent fuel from the SFP at a reference plant to dry cask storage significantly reduces the risks to public health and safety. The study states that "this study's results are consistent with earlier research studies' conclusions that spent fuel pools are robust structures that are likely to withstand severe earthquakes without leaking cooling water." The study also shows that, in the event of a radiological release, public and environmental effects are generally the same or smaller than earlier studies.

In SECY-93-127 (Reference 20), the NRC staff considered potential financial liability of a zirconium fire to determine that the overall risk at decommissioning plants does not justify the full insurance coverage once the spent fuel has sufficiently decayed. In its Staff Requirements Memorandum for SECY-93-127 (Reference 13), the Commission approved a policy that authorized reductions in commercial liability insurance coverage through the exemption process after the spent fuel had undergone an appropriate period of cooling, which the NRC staff defined as when the spent fuel could be air-cooled if the spent fuel pool was drained of water.

In NUREG/CR-6451 "A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants" (Reference 21) the representative PWR was shown to be able to be air cooled. HDI has compiled data comparing the input parameters between this representative generic analysis and like data for the IP2 and IP3. This information is provided in Table 1.

Enclosure HDI-IPEC-22-025 Page 10 of 20 Table 1 - Fuel Assembly Parameters

Parameter NUREG/CR-6451 IP2 IP3 Power 1130 MWe (§3.1.1) 3216 MWt 3188.4 MWt (Note 2) 3330 MWt (Note 1)

Assemblies 193 (§3.1.1) 193 193 MWt/Assembly 17.25 [ 3330/193] 16.66 (calculated) 16.52 (calculated)

Fuel Design 17 x 17 (p. 3-5) 15 x 15 15 x 15 Burnup 60 GWd/MTU (§3.1.1) 58.832 GWd/MTU 58.832 GWd/MTU Decay Time 17 months (§3.1.3) 28.5 months 16.5 months (519 days) (Note 2) 870 days 504 days (28.5 / 12) x (366) (16.5 / 12) x (366)

(calculated) (calculated)

Cladding Oxidation 565°C (§3.1.3) 565°C 565°C Temperature Limit MTU/Fuel Assembly 0.461 (Note 3) 0.456848 0.456848 Operating Time 1604.5 days (Note 4) 1613.0 days 1627.0 days

= (58.832) x (0.456848) x = (58.832) x

((193) / (3.216)) (calculated) (0.456848) x ((193) /

(3.1884)) (calculated)

Fuel Assembly 8.426 in (Note 5) 8.426 in 8.426 in Transverse Dimension Rod Center-to-Center 0.496 in (Note 3) 0.563 in 0.563 in Pitch Rod Outside 0.374 in (Note 3) 0.422 in 0.422 in Diameter Active Fuel Height 144 in (Note 3) 144 in 144 in Overall Fuel Rod 168 in (Note 6) 152.88 in 152.88 in Length (active + inactive length)

Fuel Rods per Fuel 289 (Note 6) 204 204 Assembly Active Volume 10,223.6 in3 (Note 7) 10,223.6 in3 10,223.6 in3 (0.1675 m3) = (144 in) x (8.426 in)2 = (144 in) x (8.426 in)2 (0.1675 m3) (calculated) (0.1675 m3)

(calculated)

Power Density 103 MWt/m3 (Note 8) 99.5 MWt/m3 98.6 MWt/m3

= ((3216) / (193)) / (0.1675) = ((3188.4) / (193)) /

(calculated) (0.1675)

(calculated)

Notes:

Enclosure HDI-IPEC-22-025 Page 11 of 20

1) The thermal power of the representative PWR in NUREG/CR -6451 is not provided; therefore, an approximate value based on 34 percent thermal efficiency is used.

2) IP3 operating power is 3188.4 MWt, and is used in Table 1 calculations. Licensed power is 3216 MWt.

3) The number of days is computed as (17 / 12) 366 days since a longer decay period results in less decay heat and is more conservative for comparing to IP3 with a shorter decay period.

4) Refer to Table 2.2 of NUREG/CR-6441 (Reference 22).

5) Operating time (0) is computed using the burnup (GWd/MTU), plant thermal power (MWt), number of in the core fuel assemblies (FA), and uranium mass per fuel assembly (MTU/FA).

= 1000 0 1

6) The PWR example in NUREG/CR-6441 utilizes a 17 17 fuel assembly from a 193-fuel assembly core with a uranium mass of 0.461 MTU.

7) Refer to Table A.3 of NUREG/CR-6441 (Reference 22).

8) Active Volume = Active Fuel Height (Fuel Assembly Transverse Dimension)2

9) Power Density = (MWt/Fuel Assembly) / Active Volume

Table 2 - Spent Fuel Rack Parameters

Parameter NUREG/CR-6451 IP2 (Note 1) IP3 (Note 1)

Design HD (Notes 2 & 3) Region 1 Flux Trap Region 1 Flux Trap Region 2 HD Region 2 HD Material SS (Note 3) Region 1 SS Region 1 SS Region 2 SS Region 2 SS Pitch (Note 4) 10.40 in Region 1 10.545 in (N/S) Region 1 10.760 in 10.765 in (E/W) Region 2 9.075 in Region 2 9.04 in Bottom Orifice 5 in dia. Regions 1 & 2 6 in dia. 0.5 in Plate 3.5 in dia.

Pedestal 5 in dia. 0.75 in Plate 3.5 in dia.

6 in x 2 in (Note 5)

Opening per Cell 8.75 in x 8.75 in Region 1 8.75 in x 8.75 in Region 1 8.83 in Region 2 8.8 in x 8.8 in Region 2 8.83 in Cell Wall Thickness 0.185 in Region 1 0.075 in Region 1 0.085 in Region 2 0.075 in Region 2 0.085 in Neutron Absorber Not mentioned Boraflex (Note 6) Boral Downcomer Width (Note 7) 3 in 1.25 in to 7 in 3.75 in to 8.86 in Plenum Height Under Racks 6 in Regions 1 & 2 8.75 in Regions 1 & 2 7.125 in (Note 8)

SFP Perimeter (Note 9) 119 ft 137 ft 4 in 137 ft 10 in Downcomer Area 29.75 ft2 (2.764 m2) 67.69 ft2 (6.288 m2) 69.05 ft2 (6.415 m2)

Fuel Assemblies in SFP 1460 1374 available locations 1345 available locations (Note 10) 990 assemblies 1151 assemblies

Notes:

1) IP2 SFP and IP3 SFP each have two fuel storage rack regions (Region 1 and Region 2).

2) Region 1 fuel storage rack design incorporates additional water spaces between cells (Flux Trap) to increase effectiveness of neutron absorber material.

3) Abbreviations: HD = high density; SS = stainless steel Enclosure HDI-IPEC-22-025 Page 12 of 20

4) Pitch = center to center distance from cell to cell

5) Cross section of each 0.75 in lifting plate (4 per rack) is combination of circle and rectangle. First dimension (3.5 in) is diameter of circle, second and third dimensions (6 in, 2 in) are length and width of rectangle.

6) Boraflex neutron absorber panels are installed in IP2 fuel racks but are not credited in SFP criticality analysis.

7) Width of downcomer varies depending on local geometry and proximity of SFP racks to wall.

8) Dimensions for IP2 are total for pedestal and bearing pad. Nominal total is given for IP3.

9) SFP perimeter is sum of wall lengths of actual SFP.

10) Fuel assemblies in SFP are as of 08/25/2020 and subject to change.

A comparison of the fuel design parameters for fuel assembly power, power density, and hydraulic resistance of the 15 x 15 fuel assemblies from IP2 and IP3 found that they are the same or conservative when compared to those for the 17 x 17 fuel assemblies modeled in NUREG/CR-6451. It can therefore be concluded that the analytical results for the NUREG/CR-6451 model fuel assembly can be conservatively applied to the IP2 and IP3 fuel assemblies.

The NUREG/CR-6451 SFP storage rack design and configuration were also compared to those for the IP2 and IP3 SFPs. Based on this comparison, it was found that the IP2 and IP3 fuel storage rack Region 2 cell pitch and IP3 cell bottom orifice dimensions are smaller than the values modeled in the NUREG. However, these differences are considered to be conservatively offset by the lower hydraulic resistance and power density of the IP2 and IP3 fuel assemblies, substantially larger SFPs and downcomer areas for improved buoyancy driven air flow and natural circulation, and the fewer number of fuel assemblies that can be stored in the fuel racks as compared to the NUREG/CR-6451 PWR model.

An IPEC analysis (Reference 23) demonstrates successful completion of the Enhanced Seismic Checklist provided in Attachment 1 to Appendix 2B of NUREG-1738 (Reference 15) for the IP2 and IP3 SFPs. Based on the analysis ( Reference 23) there is a high confidence in a low probability of failure (HCLPF) for seismic ground motions up to 1.2 g peak spectral acceleration (or with peak ground acceleration (PGA) of approximately 0.5 g), which in turn assures that the frequency of fuel uncovery from seismic events for IP2 and IP3 is less than or equal to 1x10-6 per year (Reference 23).

Conclusion

Using the Holtec Spent Fuel Pool Heat Up Calculation Methodology, the analysis submitted to NRC in Reference 4 for the IP2 and IP3 SFPs demonstrate that a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> is available before the fuel cladding temperature of the hottest fuel assembly in the SFP reaches the zirconium fire temperature of 900 degrees Celsius (°C) with a beyond design basis complete loss of SFP water inventory. In addition, the IP2 and IP3 spent fuel and SFP conditions were determined to be bounded by the NUREG/CR-6451 benchmark. Thus, demonstrating that spent fuel would be air coolable at 15 months after permanent shutdown. Regarding the dose assessments, as described above, the dose for the FHA or the BDBE SFP drain down event would be below regulatory limits.

VI. JUSTIFICATION FOR EXEMPTION AND SPECIAL CIRCUMSTANCES

As stated in 10 CFR 50.12, the Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of this part which are authorized by law, will not present an undue risk to the public health and safety, Enclosure HDI-IPEC-22-025 Page 13 of 20 and are consistent with the common defense and security. It is also stated that the Commission will not consider granting an exemption unless special circumstances are present.

As discussed below, this exemption request satisfies the provisions of 10 CFR 50.12.

A. The exemption is authorized by law

In accordance with 10 CFR 50.12, the NRC is allowed to grant exemptions from the requirements of 10 CFR Part 50 that the Commission determines are authorized by law.

The proposed reduction in onsite property damage insurance coverage to a level of $50 million is consistent with SECY-96-256 and the exemption would not result in a violation of the Atomic Energy Act of 1954, as amended, or the Commission's regulations. The NRC has granted exemptions to other licensees for insurance reductions of the same regulation being requested here by HDI and have been previously determined to be authorized by law and granted (see Section VI of this Enclosure).

B. The exemption will not present an undue risk to public health and safety

The onsite property damage insurance requirements of 10 CFR 50.54(w)(1) were established to provide financial assurance that following a significant nuclear incident, onsite conditions could be stabilized and the site decontaminated. The requirements of 10 CFR 50.54(w)(1) and the existing level of onsite insurance coverage are predicated on the assumption that the reactor is operating. With the permanent shutdown and defuel of the three reactors (IP1, IP2, and IP3) located at the IPEC site, there is a significant reduction in the potential for and severity of potential accidents, and correspondingly, a significant reduction in the potential for and severity of onsite property damage. The proposed reduction in the amount of onsite insurance coverage does not adversely affect the probability or consequences of potential accidents. The proposed level of insurance coverage is commensurate with the reduced consequences of potential nuclear accidents at the IPEC site. Therefore, granting the requested exemption will not present an undue risk to the health and safety of the public.

C. The exemption is consistent with the common defense and security

The proposed exemption would not eliminate any requirements associated with physical protection of the IPEC site and would not adversely affect HDI's ability to physically secure the site or protect special material. Physical security measures at IPEC are not affected by the requested exemption. Therefore, the proposed exemption is consistent with the common defense and security.

D. Special Circumstances

In accordance with 10 CFR 50.12(a)(2), the NRC will not consider granting an exemption to its regulations unless special circumstances are present. HDI has determined that special circumstances are present because all three reactors located at the IPEC site are permanently shutdown and defueled and the radiological source term at the site is reduced from that associated with reactor power operation. With all three IPEC power reactors permanently shutdown and defueled, the DBAs and transients postulated to occur during reactor operation are longer be possible. In particular, the potential for a release of a large Enclosure HDI-IPEC-22-025 Page 14 of 20 radiological source term to the environment from the high pressures and temperatures associated with reactor operation no longer exist.

1. Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule. (10 CFR 50.12(a)(2)(ii))

The underlying purpose of 10 CFR 50.54(w)(1) is to require sufficient property damage insurance to ensure funding of onsite post-accident recovery, stabilization, and decontamination costs following an accident at an operating nuclear power plant. The requirements of 10 CFR 50.54(w)(1) were developed taking into consideration the risks associated with the operation of an operating nuclear power reactor, including the potential consequences of a release of radioactive material from the reactor. However, the regulation does not take into consideration the reduced potential for, and consequences of, nuclear incidents at facilities that have been permanently shutdown.

The radiological consequences of accidents that remain possible at the IPEC site with all three nuclear power reactors in the permanently defueled condition are substantially lower than if the reactors were operating. After a period of 30 days following the permanent shutdown of IP3, it is no longer possible for the radiological consequences of the remaining DBAs for the IPEC nuclear power reactors to exceed the limits of the EPA PAGs at the EAB.

The proposed reduction in the level of onsite insurance coverage from $1.06 billion to

$50 million would continue to serve the underlying purpose of the 10 CFR 50.54(w)(1 )

rule by requiring a level of financial protection commensurate with the significant reduction in the probability and consequences of nuclear incidents that could occur at the IPEC site. Consistent with the NRC's conclusions documented in SECY-00-145 (Reference 7), the proposed reduction in the level of onsite insurance coverage would continue to require sufficient property damage insurance to ensure funding for onsite post-accident recovery, stabilization, and decontamination costs in the unlikely event of an accident at the IPEC site.

Therefore, application of the requirement in 10 CFR 50.54(w)(1) to maintain $1.06 billion in onsite insurance coverage is not necessary to achieve the underlying purpose of the 10 CFR 50.54(w)(1) rule and special circumstances are present as defined in 10 CFR 50.12(a)(2)(ii).

2. Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted. (10 CFR 50.12(a)(2)(iii))

Continued application of the requirement to maintain $1.06 billion in onsite insurance coverage for the IPEC site would result in undue hardship and costs being incurred by the IP1, IP2, and IP3 decommissioning trust funds for the purchase of unnecessary levels of onsite insurance coverage.

As stated in Section VII of this Enclosure, other licensees of permanently shutdown power reactors have been granted exemptions from 10 CFR 50.54(w)(1) by the NRC in the same or lower insurance amounts requested by HDI in this exemption request for the IPEC site.

Enclosure HDI-IPEC-22-025 Page 15 of 20

Therefore, compliance with the 10 CFR 50.54(w)(1) rule would result in an undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated and the special circumstances required by 10 CFR 50.12(a)(2)(iii) exist.

VII. PRECEDENT

The exemption request for 10 CFR 50.54(w)(1) is consistent with exemption requests that recently have been issued by the NRC for other nuclear power reactor facilities beginning decommissioning. Specifically, the NRC granted similar exemptions to Holtec D ecommissioning International, LLC, for Pilgrim Nuclear Power Station (Reference 26); Exelon Generation Company, LLC, for Oyster Creek (Reference 27); Southern California Edison Company, for San Onofre Nuclear Generating Station, Units 1, 2, and 3 (Reference 28); Entergy Nuclear Operations, Inc., for Vermont Yankee (Reference 29); Duke Energy Florida, Inc., for Crystal River, Unit 3 (Reference 30); Dominion Energy Kewaunee, Inc., for Kewaunee Power Station (Reference 31),

and Florida Power & Light for NextEra Energy Duane Arnold (Reference 32).

Similar to the current request, these precedents each resulted in exemptions from the requirements in 10 CFR 50.54(w)(1).

VIII.ENVIRONMENTAL ASSESSMENT

The proposed exemption meets the eligibility criterion of categorical exclusion set forth in 10 CFR 51.22(c)(25) because the proposed exemption involves: (i) no significant hazards consideration; (ii) no significant change in the types or significant increase in the amounts of any effluents that may be released offsite; (iii) no significant increase in individual or cumulative public or occupational radiation exposure; (iv) no significant construction impact; (v) no significant increase in the potential for or consequences from radiological accidents; and (vi) the requirements from which the exemption is sought involve surety, insurance, or indemnity requirements. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed exemption.

(i) No Significant Hazards Consideration Determination

HDI has evaluated the proposed exemption from 10 CFR 50.54(w)(1) for IPEC to determine whether or not a significant hazards consideration is involved by focusing on the three standards set forth in 10 CFR 50.92 as discussed below:

1. Does the proposed exemption involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed exemption has no effect on structures, systems, and components (SSCs) and is unrelated to the capability of any plant SSC to perform its design function. The proposed exemption would not increase the likelihood of the malfunction of any plant SSC.

When the exemption becomes effective, there will be no credible events that would result in doses to the public beyond the exclusion area boundary (EAB) that would exceed the Environmental Protection Agency (EPA) Protective Action Guidelines Enclosure HDI-IPEC-22-025 Page 16 of 20 (PAGs). The probability of occurrence of previously evaluated accidents is not increased, since most previously analyzed accidents will no longer be possible and the probability and consequences of the remaining design basis accidents (DBAs) are not adversely affected by the proposed exemption.

Therefore, the proposed exemption does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed exemption create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed exemption does not involve a physical alteration of the IPEC plants.

No new or different type of equipment will be installed and there are no physical modifications to existing equipment associated with the proposed exemption.

Similarly, the proposed exemption will not physically change any SSCs involved in the mitigation of any accidents. Thus, no new initiators or precursors of a new or different kind of accident are created. Furthermore, the proposed exemption does not create the possibility of a new accident as a result of new failure modes associated with any equipment or personnel failures. No changes are being made to parameters within which the plants are normally operated, or in setpoints which initiate protective or mitigative actions, and no new failure modes are being introduced.

Therefore, the proposed exemption does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed exemption involve a significant reduction in a margin of safety?

The proposed exemption does not alter the design basis or any safety limits for the IPEC plants. The proposed exemption does not impact facility operation or any plant SSC that is relied upon for accident mitigation.

Therefore, the proposed exemption does not involve a significant reduction in a margin of safety.

Based on the above, HDI concludes that the proposed exemption presents no significant hazards consideration, and, accordingly, a finding of "no significant hazards consideration" is justified.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

There are no expected changes in the types, characteristics, or quantities of effluents discharged to the environment associated with the proposed exemption. There are no materials or chemicals introduced into the IPEC plants that could affect the characteristics of types of effluents released offsite. In addition, the method of operation of waste processing systems will not be affected by the exemption. The proposed exemption will not result in changes to the design basis requirements of SSCs that function to limit or monitor the release of effluents. All the SSCs associated with limiting the release of effluents will continue to be able to perform their functions.

Enclosure HDI-IPEC-22-025 Page 17 of 20 Therefore, the proposed exemption will result in no significant change to the types or significant increase in the amounts of any effluents that may be released offsite.

(iii) There is no significant increase in individual or cumulative public or occupational radiation exposure.

The proposed exemption does not involve any physical alterations to the configuration of the plants or any changes to operation of the facility that could lead to a significant increase in individual or cumulative occupational radiation exposure to either the workforce or the public.

(iv) There is no significant construction impact.

No construction activities are associated with the proposed exemption.

(v) There is no significant increase in the potential for or consequences from radiological accidents.

See the no significant hazards considerations discussion in Item (i)(1) above.

(vi) The requirements from which exemption is sought involve surety, insurance or indemnity requirements.

The requirements from which the exemption is sought involve financial protection and for the indemnification and limitation of liability in accordance with Section 170 of the Atomic Energy Action of 1954, as amended, and 10 CFR 50.54(w)(1).

IX. CONCLUSION

In accordance with 10 CFR 50.12, HDI is requesting a permanent exemption from 10 CFR 50.54(w)(1) for the IPEC site. Based on the considerations discussed above, the requested exemption is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security. In addition, special circumstances are present as set forth in 10 CFR 50.12.

REFERENCES

1. Entergy letter to U.S. NRC, "Notification of Unit 1 Transfer of 160 Spent Fuel Assemblies from the Spent Fuel Pool to the Indian Point Independent Spent Fuel Storage Installation," (Accession No. ML083510667), dated December 11, 2008

2. Entergy letter to U.S. NRC, "Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel, Indian Point Nuclear Generating Unit No. 2," (Accession No. ML20133J902), dated May 12, 2020

3. Entergy letter to U.S. NRC, "Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel, Indian Point Nuclear Generating Unit No. 3," (Accession No. ML21131A157), dated May 11, 2021

4. Holtec Decommissioning International, LLC (HDI) letter to U.S. NRC, "Supplement to HDI Request for Exemptions from Certain Emergency Planning Requirements of Enclosure HDI-IPEC-22-025 Page 18 of 20 10 CFR 50.47 and 10 CFR 50, Appendix E, Indian Point Nuclear Generating Unit Nos. 1, 2, and 3 Including Site-Specific Calculations," (Accession No. ML22032A017), dated February 1, 2022

5. U.S. NRC Commission Paper, SECY-96-256, "Changes to the Financial Protection Requirements for Permanently Shutdown Nuclear Power Reactors, 10 CFR 50.54(w) and 10 CFR 140.11, (Accession No. ML15062A483), dated December 17, 1996

6. Staff Requirements Memorandum, "SECY-96-256, Changes to Financial Protection Requirements for Permanently Shutdown Nuclear Power Reactors," (Accession No. ML15062A454), dated January 28, 1997

7. U.S. NRC Commission Paper, SECY-00-145, "Integrated Rulemaking Plan for Nuclear Power Plant Decommissioning," (Accession No. ML003721626), dated June 28, 2000

8. U.S. NRC Commission Paper, SECY-01-0100, "Policy Issues Related to Safeguards, Insurance, and Emergency Preparedness Regulations at Decommissioning Nuclear Power Plants Storing Fuel in Spent Fuel Pools," (Accession No. ML011450420), dated June 4, 2001

9. Holtec Decommissioning International, LLC (HDI) letter to U.S. NRC, Revision to Holtec Decommissioning International, LLC (HDI) Request for Exemptions from Certain Emergency Planning Requirements of 10 CFR 50.47 and 10 CFR Part 59, Appendix E, (Accession No. ML22033A348) dated February 2, 2022

10.U.S. NRC letter to Entergy, "Indian Point Nuclear Generating Unit No. 2 - Issuance of Amendment No. 294 RE: Permanently Defueled Technical Specifications (EPID L-2019-LLA-0079)," (Accession No. ML20081J402), dated April 28, 2020

11.U.S. NRC letter to Entergy, "Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment No. 270 Re: Permanently Defueled Technical Specifications (EPID L-2020- LLA-0090)," (Accession No. ML21074A000), dated April 22, 2021

12.Calculation IP-CALC-19-00003, "Post-Permanent Shutdown Analyses of Fuel Handling, Waste Handling, and High Integrity Container Drop Accidents for Indian Point Units 2 and 3," Revision 0, March 13, 2019

13.U.S. Nuclear Regulatory Commission Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, (ADAMS Accession No. ML003716792), dated July 2000

14.U.S. NRC Commission Paper, SECY-99-168, "Improving Decommissioning Regulations for Nuclear Power Plants," (ADAMS Accession No. ML992800087), dated June 30, 1999

15.U.S. NRC NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," (Accession No. ML010430066), dated February 2001 Enclosure HDI-IPEC-22-025 Page 19 of 20 16.U.S. NRC NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," (Accession Nos. ML023470327 and ML023500228), dated October 2002

17.Calculation IP-CALC-18-00066, "Shine Dose to Exclusion Area Boundary and Control Room from Spent Fuel Pool During SAFSTOR," Revision 1, July 15, 2019

18.U.S. Environmental Protection Agency, "Protective Action Guides and Planning Guidance for Radiological Incidents," EPA-400/R-17-001, (EPA PAG Manual), dated January 2017

19.U.S. NRC NUREG-2161, "Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor," (Accession No. ML14255A365), dated September 2014

20. SECY-93 -127, "Financial Protection Required of Licensees of Large Nuclear Power Plants During Decommissioning," (Accession No. ML12257A628), dated May 10, 1993

21. U.S. NRC NUREG/CR-6451, "A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants," (Accession No. ML082260098), dated April 1997

22.U.S. NRC NUREG/CR-6441 (BNL-NUREG-52494), "Analysis of Spent Fuel Heatup Following Loss of Water in a Spent Fuel Pool," (Accession No. ML021050336), dated March 2002

23.IP-CALC-2 0-00022, "Evaluation of Unit 2 & Unit 3 Spent Fuel Pool (SFP) per NUREG-1738 Appendix B Seismic Checklist," Revision 0

24.Letter from U.S. NRC to Holtec Decommissioning International, LLC, "Pilgrim Nuclear Power Station - Exemption from the Requirements of 10 CFR 50.54(w)(1) Concerning Onsite Property Damage Insurance (EPID L-2019-LL E-0006)," (Accession No. ML19281D742), dated January 6, 2020

25.Letter from U.S. NRC to Exelon Generation Company, LLC, "Oyster Creek Nuclear Generating Station - Exemption from the Requirements of 10 CFR 50.54(w)(1),

Concerning Onsite Property Damage Insurance (EPID L-2018-LLE-0004)," (Accession No. ML18228A852), dated December 19, 2018

26.U.S. NRC Exemption issuance to Southern California Edison Company, "San Onofre Nuclear Generating Station, Units 1, 2, and 3 - Exemption from the Requirements in Section 54(w)(1) of Title 10 of the Code of Federal Regulations (10 CFR), Concerning Onsite Property Damage Insurance (CAC Nos. L53097, L53098, and L53099)," (Accession No. ML17355A023), dated January 5, 2017

27. Letter from U.S. NRC to Entergy Nuclear Operations, Inc., "Vermont Yankee Nuclear Power Station - Exemption from the Requirements of Title 10 of the Code of Federal Regulations, Part 50, Section 50.54(w)(1) Concerning Insurance for Post-Accident Site Decontamination (CAC No. MF3981)," (Accession Nos. ML16012A193 and ML16012A197), dated April 15, 2016

28. Letter from U.S. NRC to Crystal River Nuclear Plant, "Crystal River Unit 3 Nuclear Generating Plant - Exemption from the Requirements of Title 10 of the Code of Federal Regulations, Section 50.54(w)(1) Concerning Insurance for Post-Accident Enclosure HDI-IPEC-22-025 Page 20 of 20 Site Decontamination (TAC No. L53108)," (Accession Nos. ML16020A448 and ML16020A463), dated March 16, 2016

29.Letter from U.S. NRC to Dominion Energy Kewaunee, Inc., "Kewaunee Power Station -

Exemption from the Requirements of Title 10 of the Code of Federal Regulations, Part 50, Section 50.54(w)(1) Concerning Insurance for Post-Accident Site Decontamination (TAC No. MF3915)," (Accession No. ML15033A245), dated April 3, 2015

30.Letter from U.S. NRC to Florida Power & Light Company, Duane Arnold Energy Center

-Exemption from the Requirements of 10 CFR 50.54(w)(1) Concerning Onsite Property Damage Insurance (EPID L-2020-LL E-01 14), (Accession Nos MLl21068A197 and ML21068A211), dated May 11, 2021