NG-20-0035, Request for Exemption from 10 CFR 50.54(w)(1)

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

ML20198M579
Person / Time
Site:	Duane Arnold
Issue date:	07/16/2020
From:	Dean Curtland NextEra Energy Duane Arnold
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NG-20-0035
Download: ML20198M579 (17)

Text

Duane Arnold Energy Ce nter 32 77 DAEC Road Palo, Iowa 52324 July 16, 2020 NG-20-0035 10 CFR 50.12 10 CFR 50.54(w)(1)

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington , DC 20555-0001 Duane Arnold Energy Center Docket No. 50-331 Renewed Op. License No. DPR-49 Request for Exemption from 10 CFR 50.54(w)(1)

Reference:

Letter from NEDA (D. Curtland) to USN RC, "Request for Exemption from Portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E," dated April 2, 2020 (ML20101M779)

Pursuant to 10 CFR 50.12, NextEra Energy Duane Arnold , LLC (NEDA) requests exemption from 10 CFR 50.54(w)(1) for the licensed owners of the Duane Arnold Energy Center (DAEC) .

10 CFR 50.54(w)(1) 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 ."

NEDA is requesting an exemption to 10 CFR 50.54(w)(1) to reduce the DAEC minimum onsite insurance coverage to $50 million.

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. However, the regulation does not take into consideration the reduced potential for, and consequences of, such nuclear incidents at permanently shut down facilities .

By letter dated March 2, 2020 (ML20062E489), pursuant to 10 CFR 50.82(a)(1 )(i), NEDA submitted a certification to the NRC indicating its intention to permanently cease power operations at DAEC on October 30, 2020 . After the certifications of permanent cessation of power operation and of permanent removal of fuel from the reactor vessel are docketed for DAEC, in accordance with 10 CFR 50.82(a)(1)(i) and (ii) , and pursuant to 10 CFR 50.82(a)(2) ,

NG-20-0035 Page 2 of 2 the 10 CFR 50 license will no longer authorize reactor operation or emplacement or retention of fuel in the reactor vessel. Consequently, no additional fission products will be generated from the plant after shutdown and the decay heat load on the spent fuel will continue to decline. The proposed exemption would allow a reduction in onsite insurance coverage to a level that is commensurate with the future of the facility and the underlying purpose of the rule.

NEDA evaluated the effects of an accident where the spent fuel assemblies in the Spent Fuel Pool (SFP) are uncovered following a drain down event. The analysis submitted in the Referenced letter demonstrates that a significant release of radioactive material from the spent fuel is not possible after 10 months following permanent cessation of power operations.

Based on the length of time it would take for the adiabatic heat up to occur, there is ample time to respond to any partial drain down event that might cause such an occurrence by restoring SFP cooling or makeup, or providing SFP spray. As a result, the likelihood that such a scenario would progress to a zirconium fire is deemed not credible.

The requested exemptions are permissible under 10 CFR 50.12 because they are authorized by law, will not present an undue risk to the public health and safety, are consistent with the common defense and security, and present special circumstances .

NEDA requests review and approval of this exemption request by July 1, 2021 with an effective date no less than 10 months after shutdown .

If you have questions regarding this submittal, please contact J. Michael Davis, Licensing Manager at 319-851-7032.

l!kt001 ~

Dean Curtland Site Director, Duane Arnold Energy Center NextEra Energy Duane Arnold, LLC

Enclosure:

Request for Exemption from 10 CFR 50.54(w)(1) cc: NRC Region lllAdministrator NRC Resident Inspector NRC Project Manager A. Leek, State of Iowa

NG-20-0035 Enclosure Duane Arnold Energy Center Request for Exemption from 10 CFR 50.54(w)(1) 14 pages follow

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 1 of 14 Summary Description Pursuant to 10 CFR 50.12, NextEra Energy Duane Arnold, LLC (NEDA) requests exemption from 10 CFR 50.54(w)(1) for the Duane Arnold Energy Center (DAEC) on behalf of each of the licensed owners of the DAEC. 10 CFR 50.54(w)(1) 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. NEDA is requesting an exemption from 10 CFR 50.54(w)(1) to reduce the DAEC minimum onsite insurance coverage to $50 million.

II Background DAEC is located near the town of Palo, Iowa in Linn County, and consists of approximately 500 acres adjacent to the Cedar River. DAEC is a boiling water reactor with a rated thermal power of 1912 MWt. A detailed description of the plant is given in the DAEC Updated Final Safety Analysis Report (UFSAR). An Independent Spent Fuel Storage Installation (ISFSI) is situated on the owner controlled area. A detailed description of the ISFSI is found in the "Updated Final Safety Analysis Report for the Standardized Nuhoms Horizontal Modular Storage System for Irradiated Nuclear Fuel."

By letter dated March 2, 2020 (Reference 1), pursuant to 10 CFR 50.82(a)(1 )(i), NEDA submitted a certification to the NRC indicating its intention to permanently cease power operations at DAEC on October 30, 2020. Once fuel has been permanently removed from the reactor vessel, NEDA will submit a written certification to the NRC, in accordance with 10 CFR 50.82(a)(1)(ii) that meets the requirements of 10 CFR 50.4(b)(9). Upon docketing of these certifications, the 10 CFR Part 50 license for DAEC will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2).

Chapter 15 of the UFSAR describes the safety analysis aspects of DAEC that were evaluated to demonstrate that the plant could be operated safely and that radiological consequences from postulated accidents do not exceed regulatory requirements. When the reactor is permanently defueled, the SFP and its supporting systems will be dedicated only to spent fuel storage.

Irradiated fuel will continue to be stored in the SFP and the ISFSI until it is removed by the Department of Energy. Additionally, the reactor vessel assembly and supporting structures and systems are no longer in operation and have no function related to the safe storage and management of irradiated fuel in the SFP. Consequently, the only UFSAR Chapter 15 design-basis accident scenario that remains credible in the permanently defueled condition, with fuel stored in the SFP, is a fuel handling accident (FHA).

When the reactor is permanently defueled, the SFP and its supporting systems will be modified and dedicated only to spent fuel storage. A SFP cooling and clean-up system is provided to remove decay heat from the spent fuel stored in the SFP and to maintain a specified water temperature, purity, clarity and level.

Ill Detailed Description Pursuant to 10 CFR 50.12, NextEra Energy Duane Arnold, LLC (NEDA) requests exemption from 10 CFR 50.54(w)(1) on behalf of each of the licensed owners of the DAEC. NEDA is requesting an exemption to 10 CFR 50.54(w)(1) to reduce the minimum coverage limit to

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 2 of 14

$50 million for DAEC. 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 NRG 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 within 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.

IV Discussion 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.

This regulation does not take into consideration the reduced potential for, and consequences of, such nuclear incidents at permanently shut down facilities. DAEC is a single reactor site and the reactor will be permanently shut down and defueled. The proposed exemption would allow a reduction in the level of onsite insurance coverage to a level that is commensurate with the planned permanently defueled status of DAEC and the underlying purpose of the 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 a permanently shut down and defueled reactor, nuclear accidents involving the reactor and its associated systems, structures and components 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 the volatile components like iodine and noble gasses, decay away; thus reducing the inventory of radioactive materials available for release.

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 3 of 14 Although the potential for, and consequences of, nuclear accidents decline substantially after a plant permanently defuels its reactor, 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 a permanently shut down and defueled reactor may contain an inventory of radioactive liquids, activated reactor components, and contaminated materials. For purposes of modifying the amount of onsite insurance coverage maintained by a permanently shut down and defueled reactor licensee, the potential radiological consequences of these non-operating reactor nuclear incidents are appropriate to consider, despite their very low probability of occurrence.

NRC Proposed Rulemaking 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 shut down. The results of these evaluations were summarized in SECY-96-256 (Reference 2) and the NRC staff recommended course of action was approved by the Commission in a Staff Requirements Memo (SRM). These documents established the basis for the NRC exercising its discretionary authority to specify an appropriate level of onsite insurance coverage for permanently shut down nuclear power reactors.

In SECY-97-186 (Reference 3), the NRC staff proposed rulemaking for Commission approval that was consistent with SECY-96-256, Option 2. In SECY-97-186, the NRC staff proposed changes to 10 CFR 50.54(w)(1) that would establish appropriate levels of onsite insurance coverage for plants that are permanently shut down and defueled and that meet specified facility configurations during permanent shutdown.

In October 1997, the NRC published a proposed rulemaking to amend regulations governing liability coverage for permanently shut down nuclear plants. The proposed rulemaking established four different configurations for permanently shut down plants that encompassed anticipated spent fuel characteristics and storage modes during the period between permanent shutdown and termination of the license. The rulemaking proposed financial protection requirements for each of the four specified plant configurations, including a configuration where the plant is permanently shut down, the reactor defueled, and the spent fuel stored in the spent fuel pool is not susceptible to a zircaloy cladding failure or gap release caused by an incipient fuel cladding failure if the pool is accidentally drained.

However, the NRC staff rulemaking efforts were suspended prior to issuing the final rule when it was realized that an NRC staff-approved technical basis did not exist for generic decay times after which the zirconium cladding failure concern could be eliminated. The proposed changes to regulations governing onsite insurance coverage were subsequently included in a risk-informed, integrated rulemaking initiative for decommissioning nuclear power plants, which has yet to be acted on. This rulemaking initiative, documented in SECY-00-145 (Reference 4),

included onsite insurance coverage requirements based on the proposed decommissioning insurance rulemaking issued in October 1997, as modified to address the public comments received in response to that proposed rulemaking. The modified rulemaking, as incorporated into SECY-00-145, would have allowed the minimum onsite insurance coverage to be reduced to $25 million once the spent fuel in the SFP is no longer thermal-hydraulically capable of sustaining a zirconium fire, based on a plant-specific analysis.

As discussed in the staff response to a question in SECY-00-145 (see "NRC Staff Responses to NEI White Paper Comments on Improving Decommissioning Regulations," response to

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 4 of 14 Question 3):

"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."

In addition, as discussed in the staff response to a question in SECY-00-145 (see "NRC Staff Responses to NEI White Paper Comments on Improving Decommissioning Regulations, in response to Question 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 a memorandum dated August 16, 2002, the NRC Executive Director for Operations provided the NRC Commissioners a status of the regulatory exemptions for plants in decommissioning.

This memorandum stated:

"In the absence of any anticipated nuclear power plant decommissionings in the near term, the staff believes that there is no immediate need for moving forward with a majority of the decommissioning regulatory improvement work that is currently planned.

Specifically, broad scope regulatory improvements for decommissioning nuclear power plants do not appear to be of sufficient priority given a lack of future licensees that would benefit. at this time. Due to other higher priorities, resources are being deferred for decommissioning rulemakings that are not currently in progress or not related to security .... If any plants do unexpectedly shutdown permanently, decommissioning regulatory issues would continue to be addressed through the exemption process in a manner similar to the current practice."

Thus, the proposed rulemaking process changes for decommissioning plants discussed above were stopped in deference to the exemption process that had been used for previous licensees.

In January 2018, NRC issued its "Regulatory Analysis for Regulatory Basis: Regulatory Improvements for Power Reactors Transitioning to Decommissioning" NRC-2015-0070, RIN 3150-AJ59 (Reference 5). In Section 5.8 of this Regulatory Basis document, the NRC staff assessed offsite and onsite financial protection requirements and indemnity agreements and proposed alternatives that include an alternative "FP-2." This alternative would involve rulemaking to reduce the on-site property damage requirement in 10 CFR 50.54(w) to $50 million for a reactor that "is defueled and permanently shut down, and spent fuel in the SFP has decayed and cooled sufficiently that it cannot heat up to clad ignition temperature within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> under adiabatic conditions." In Section 8.2.9 of the Regulatory basis document, the NRC staff has recommended alternative FP-2.

V Technical Evaluation Accident Analysis Overview 10 CFR 50.82(a)(2) specifies that the 10 CFR Part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel after docketing the certifications for permanent cessation of operations and permanent removal of fuel from the

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 5 of 14 reactor vessel in accordance with 10 CFR 50.82(a)(1 ). Following the termination of reactor operations at DAEC and the permanent removal of the fuel from the reactor vessel, the postulated accidents involving failure or malfunction of the reactor and supporting structures, systems and components are no longer applicable.

A summary of the postulated radiological accidents analyzed for the permanently shut down and defueled condition is presented below. Current Federal guidance provided in the EPA's, "Protective Action Guides and Planning Guidance for Radiological Incidents, EPA-400/R-17/001," (Reference 6) Section 2.2.4, "PAGs and Nuclear Facilities Emergency Planning Zones (EPZ)," states that the EPZ is based on the maximum distance at which a PAG might be exceeded.

Section 5.0 of ISG-02 (Reference 7) indicates that site-specific analyses should demonstrate that: (1) the radiological consequences of the remaining applicable postulated accidents would not exceed the limits of the EPA PAGs at the Exclusion Area Boundary (EAB); (2) in the event of a beyond design basis event resulting in the partial drain down of the SFP to the point that cooling is not effective, there is at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (assuming an adiabatic heat up) from the time that the fuel is no longer being cooled until the hottest fuel assembly reaches 900°C; (3) adequate physical security is in place to assure implementation of security strategies that protect against spent fuel sabotage; and (4) in the unlikely event of a beyond design basis event resulting from a loss of all SFP cooling, there is sufficient time to implement pre-planned mitigation measures to provide makeup or spray to the SFP before the onset of zirconium cladding ignition.

NEDA also described the applicable DAEC analyses in the "Request for Exemption from Portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E," dated April 2, 2020 (Reference 8).

Specific analyses are summarized in the following sections.

A. Consequences of Design Basis Events The postulated design basis accident that will remain applicable to DAEC in its permanently shut down and defueled condition is the FHA in the reactor building where the SFP is located.

Analysis based on the FHA was performed to determine the dose to personnel in the Control Room and to the public at the Exclusion Area Boundary (EAB or "Site Boundary") as a function of time after shutdown. The FHA analyzed used the calculated number of fuel pin failures based on a drop of the assembly into the reactor core. Dose consequences for a drop over the reactor core bound the consequences of a drop of an assembly in the SFP due to the shorter drop height equating to fewer fuel pin failures. The analysis used the Alternative Source Term methodology from Regulatory Guide 1.183, and concluded that the dose at the EAB 19 days after shutdown (with open containment) is less than 1 rem TEDE, which is below the EPA PAG threshold of 1 rem for recommended evacuation.

B. Consequences of a Beyond Design Basis Event NEDA performed an analysis, submitted as Attachment 2 to Reference 8, which compares the conditions for the hottest fuel assembly stored in the DAEC fuel pool to a criterion proposed in SECY-99-168 (Reference 9) applicable to offsite emergency response for a unit in the decommissioning process. This criterion considers the time for the hottest assembly to heat up from 30 degrees Celsius (°C) to 900°C adiabatically.

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 6 of 14 Based on the limiting fuel assembly for decay heat and adiabatic heatup analysis, at 1O months after shutdown (1 O months of decay time), the time for the hottest fuel assembly to reach 900°C is >10 hours after the assemblies have been uncovered. As stated in NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants" (February 2001) (Reference 10), 900°C is an acceptable temperature to use for assessing 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 adiabatic heatup to occur, there is ample time to respond to any partial drain down event that might cause such an occurrence by restoring cooling or makeup, or providing spray. As a result, the likelihood that such a scenario would progress to a zirconium fire is not deemed credible.

C. Consequences of Other Analyzed Events Spent Fuel Pool Drain Down The analysis in Attachment 3 of Reference 8 assumes a complete loss of SFP water inventory while in safe storage. A loss of water shielding above the fuel could increase the offsite radiation levels because of the gamma rays streaming up out of the pool being scattered back to a receptor at the site boundary. The offsite radiological impact of a postulated complete loss of SFP water was assessed. It was determined that the gamma radiation dose rate at the EAB would be less than the EPA PAG exposure levels. The extended period required to exceed the integrated PAG limit of 1 rem TEDE would allow sufficient time to develop and implement onsite mitigative actions and provide confidence that additional offsite measures could be taken without planning if efforts to reestablish shielding over the fuel are delayed. The analysis shows that after approximately 9 months (0.75 years) of decay time, the time to exceed the PAG limit of 1 rem TEDE at the EAB following a SFP drain down is approximately 198 days, or about 6.5 months. This value can be compared to the 1O hour time limit for zirconium ignition in ISG-02 mitigative actions will have been taken far in advanced of exceeding 1 rem TEDE at the EAB.

Therefore, conditions 1O months following reactor shutdown are bounded.

The dose rate to the Control Room was determined to be <0.03 mrem/hr. While there are no acceptance criteria for the Control Room in ISG-02, the dose rate values are considered reasonably low.

Consequences of a Beyond-Design Basis Earthquake NUREG-1738 (Reference 10) 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 6) at the EAB.

The risk associated with zirconium cladding fire events decreases as the spent fuel ages. As

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 7 of 14 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 11) 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 11) 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 12), 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 14) the representative BWR was shown to be able to air cool the fuel within a 7-month window. NEDA has compiled data comparing the input parameters between this representative generic analysis and like data for the DAEC. This information is provided in Table 1.

The review of the data shows that the spent fuel pool and nuclear fuel configurations both have an increased capability to remove heat from the fuel. With respect to the heat removal capability of the spent fuel pool, several points of margin compared to the Model Plant include: a 52%

reduction in freshly discharged bundles, the spent fuel pool is 75% full versus 100 % full, resulting in additional air flow and cooling pathways, and the newly discharged fuel is a 1Ox10 array with partial rods and large water rods resulting in a >25% increase in fuel clad surface area and air flow channels within the fuel. With respect to the fuel, the max assembly burn-up and power density are higher, however, these are offset by the 3-month additional source term decay time (10 months versus 7 months) and the increased surface area/heat transfer area of the 1ox10 fuel.

Based on the margins found in the increased heat removal capabilities of the spent fuel pool racks, the additional heat removal capabilities of the nuclear fuel geometry, and the 3 month additional decay beyond the 7 month source decay required for the model plant, it is concluded that reasonable assurance exists that the Duane Arnold spent fuel pool conditions are bounded by the NUREG/CR-6451 analysis, demonstrating that the Duane Arnold Spent fuel can be air cooled following 10 months of source term decay.

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 8 of 14 Table 1 - NUREG/CR-6451 Spent Fuel Zirconium Fire Comparison NUREG/CR-6451 DAEC Item Parameter Model Plant Value Difference Value (Reference) 1 Plant Data a) Power a) 3300 MWt a) 1912-MWt (UFSAR Sect. 1.1) 396 fewer newly b) Assemblies b) 764 b) 368 (UFSAR, Sect. 1.2.5.1.1) discharged assemblies c) MWUAssembly c) 4.3 c) 5.2 (Calculated) which reduces the overall heat load in the pool 2 Spent Fuel Pool Assemblies ~ssemblies DAEC fuel pool is only 75%

Storage Data full increasing the quantity of a) Total Capacity a) 3300 a) 2411 (UFSAR, Sect. 9.1.2.2.1) air flow passages to allow for b) Quantity in SFP b) 3300 b) 1818 (Physical Inventory) fuel and fuel pool rack cooling 3 SFP Rack Design a) Design a) High Density a) High Density (UFSAR, Sect. Qrifice:

b) Material b) Stainless Steel 9.1.2.2.2) Bounded by maximum fuel c) Pitch c) 6.255" b) Stainless Steel (Holtec) orifice size = 3.505" (Fuel d) Orifice Size d) 4" c) 6.06" Holtec Rack (Holtec dwg Bundle Drawings)

M453-012) d) 3.625" Holtec (Holtec dwg M453-003)

The 368 bundles removed from the reactor in November 2020 will be inserted into the HOLTEC racks in the SFP.

4 Fuel Max Assembly 40 GWD/MTU 50.4 GWD/MTU (Calculated) Potentially up to 27%

Burn up higher heat due to increased max assembly burn up 5 Source Term 7 Months 10 Months (Reference 8) 15% less decay heat Decay contribution from hot bundles due to longer source term decay time 6 Zirconium 565°C 565°C Same Oxidation Temperature Limit 7 Fuel Design 7x7 bundle design 10x10 bundle design (Calculation M19- >25% Increase heat transfer 001) surface area.14 partial rods and 2 large water rods increase air flow pathways NEDA conducted a seismic evaluation in response to Recommendation 2.1 of the Near Term Task Force (NTTF) review of the accident at the Fukushima Daiichi nuclear facility. This evaluation included the spent fuel pool and was submitted to the NRC for review (Reference 15). This evaluation provides a specific assessment of earthquake probabilities versus ground acceleration for the Duane Arnold Energy Center, and concludes, regardless of response spectral frequency, the probability is less than 2 x 1o-6/year. The NRC review of this evaluation is documented in References 16 and 17.

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 9 of 14 VI JUSTIFICATION FOR EXEMPTIONS AND SPECIAL CIRCUMSTANCES 10 CFR 50.12 states that the Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of Part 50 which are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the defense and security. 10 CFR 50.12 also states that the Commission will not consider granting an exemption unless special circumstances are present. As discussed below, this exemption request satisfies the provisions of Section 50.12.

A. The exemptions are authorized by law 10 CFR 50.12 allows the NRC to grant exemptions from the requirements of 10 CFR Part 50.

The proposed exemption would not result in a violation of the Atomic Energy Act of 1954, as amended, or the Commission's regulations. Exemptions granted to other licensees for insurance reductions of the same regulation being requested here by NEDA have been previously determined to be authorized by law and granted, see Section E below.

In addition, the requested exemption is consistent with the guidelines presented by the NRC staff in SECY-96-256. Therefore, the exemption is authorized by law.

8. The exemptions will not present an undue risk to public health and safety The requirements of 10 CFR 50.54(w)(1) and the existing level of onsite insurance coverage for DAEC are predicated on the assumption that the reactor is operating. However, DAEC will be permanently shut down on October 30, 2020 and defueled shortly thereafter. The planned permanent defueled status of the facility will result in a significant reduction in the number 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 impact the probability or consequences of potential accidents.

As discussed in Reference 8, revised radiological analyses have been developed that show that 19 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the Environmental Protection Agency (EPA) Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the SFP which show that, within 1O months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the radiological consequences of the event will not exceed the limits of the EPA Protective Action Guides at the exclusion area boundary (EAB).

Additionally, the offsite and Control Room radiological impacts of a postulated complete loss of SFP water were assessed. It was determined that the gamma radiation dose rate at the EAB would be limited to small fractions of the EPA PAG exposure levels and the dose rate in the Control Room will be below 0.03 mRem/hr.

The proposed level of insurance coverage is commensurate with the risk and reduced consequences of potential nuclear accidents at DAEC once it is permanently defueled.

Therefore, granting the requested exemption will not present an undue risk to the health and safety of the public.

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 10 of 14 C. The exemptions are consistent with the common defense and security The proposed exemption would not eliminate any requirements associated with physical protection of the site and would not adversely affect DAE C's ability to physically secure the site or protect special nuclear material. Physical security measures at DAEC are not affected by the requested exemption. Therefore, the proposed exemptions are consistent with the common defense and security.

D. Special Circumstances Pursuant to 10 CFR 50.12(a)(2), the NRC will not consider granting an exemption to its regulations unless special circumstances are present. NEDA has determined that special circumstances are present as discussed below.

Special circumstances exist at DAEC because the plant will be permanently shut down and defueled and the radiological source term at the site will be reduced from that associated with reactor power operation. With the reactor power plant permanently shut down and defueled, the design basis accidents and transients postulated to occur during reactor operation will no longer be possible. In particular, the potential for a release of a large radiological source term to the environment from the high pressures and temperatures associated with reactor operation will 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 a 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 permanently shut down facilities.

The radiological consequences of accidents that will remain possible at DAEC in the permanently shut down and defueled condition are substantially lower than those at an operating plant.

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 rule by requiring a level of financial protection commensurate with the significant reduction in the probability and consequences of nuclear incidents at DAEC. Consistent with the NRC's conclusions documented in SECY-00-145 (Reference 4), 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 DAEC. 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 this rule and special circumstances are present as defined in 10 CFR 50.12(a)(2)(ii).

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 11 of 14

2. Compliance would result in 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. (10 CFR 50.12(a)(2)(iii))

Continued application of the requirement to maintain $1.06 billion in onsite insurance coverage for DAEC would result in undue hardship and costs being incurred by the DAEC decommissioning trust fund for the purchase of unnecessary levels of onsite Insurance coverage. As discussed in Section E below, other licensees of permanently shut down power reactors have been granted exemptions by the NRC to the subject regulation in the same or lower insurance amounts being requested by NEDA for DAEC.

Therefore, compliance with the 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.

E. 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 Entergy Nuclear Operations, Inc., for Vermont Yankee (ML16012A193); to Duke Energy Florida, Inc. for Crystal River Unit 3 (ML16020A463); to Southern California Edison Company for SONGS, Units 1, 2, and 3 (ML17355A023); to Omaha Public Power District for Fort Calhoun Station, Unit 1 (ML18031A057); to Dominion Energy Kewaunee, Inc. for KPS (ML15033A245); and to Entergy Nuclear Operations, Inc., for Pilgrim (ML19281D742). Similar to the current request, these precedents each resulted in exemptions from the requirements in 10 CFR 50.54(w)(1) to reduce the minimum coverage limit of 10 CFR 50.54(w)(1) to $50 million.

VII ENVIRONMENTAL ASSESSMENT The proposed exemption meets the eligibility criterion for 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 The requested exemptions from 10 CFR 50.54(w)(1) would allow NEDA to reduce the DAEC minimum onsite insurance coverage to $50 million. NEDA has evaluated the proposed exemption to determine whether or not a significant hazards consideration is

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 12 of 14 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 no effect on 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. The proposed changes do not affect accident initiators or precursors, nor do they alter design assumptions that could increase the probability or consequences of previously evaluated accidents.

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 Guides (PAGs). The probability of occurrence of previously evaluated accidents is not increased because most previously analyzed accidents will no longer be able to occur and the probability and consequences of the remaining postulated accident, a Fuel Handling Accident, is unaffected 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 plant. No new or different type of equipment will be installed and there are no physical modifications to existing equipment associated with the proposed exemptions. 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 plant is normally operated, or in the 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 plant. The proposed exemptions do not impact station 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.

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 13 of 14 Based on the above, NEDA concludes that the proposed exemption present 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 plant that could affect the characteristics or 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. The SSCs associated with limiting the release of effluents will continue to be able to perform their functions. 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 exemption will result in no expected increases in individual or cumulative occupational radiation exposure on either the workforce or the public. There are no expected changes in normal occupational doses. Likewise, the dose of the postulated accident is not impacted by the proposed exemptions.

(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) Surety, insurance or indemnity requirements.

The requirements from which the exemption is sought involve insurance provisions in 10 CFR 50.54(w)(1 ).

Duane Arnold Energy Center Enclosure to NG-20-0035, Page 14 of 14 VIII REFERENCES

1. Letter from NEDA (M. Nazar) to USNRC, "Certification of Permanent Cessation of Power Operations," NG-19-0136, dated March 2, 2020 (ML20062E489)

2. SECY-96-256, "Changes to the Financial Protection Requirements for Permanently Shutdown Nuclear Power Reactors," dated December 17, 1996(ML15062A483)

3. SECY-97-186, "Changes to the Financial Protection Requirements for Permanently Shutdown Nuclear Power Reactors, 10 CFR 50.54(w) and 10 CFR 140.11," dated August 13, 1997 (ML992930019)

4. USNRC, "Integrated Rulemaking Plan for Nuclear Power Plant Decommissioning,"

Commission Paper SECY-00-145, June 28, 2000 (ML003721626)

5. NRC-2015-0070, RIN 3150-AJ59, "Regulatory Analysis for Regulatory Basis: Regulatory Improvements for Power Reactors Transitioning to Decommissioning," dated January 2018 (ML17332A075)

6. EPA-400/R-17/001, "PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents," dated January 2017(ML17044A073)

7. NSIR/DPR-ISG-02, "Interim Staff Guidance, Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants," dated May 11, 2015 (ML14106A057)

8. Letter from NEDA (D. Curtland) to USN RC, "Request for Exemption from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E," NG-19-0142, dated April 2, 2020 (ML20101M779)

9. Commission Paper SECY-99-168, "Improving Decommissioning Regulations for Nuclear Power Plants," dated June 30, 1999 (ML992800087)

10. NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," dated February 2001 (ML010430066)

11. 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(ML14255A365)

12. SECY-93-127, "Financial Protection Required of Licensees of Large Nuclear Power Plants During Decommissioning," dated May 10, 1993(ML12257A628)

13. Staff Requirements Memorandum for SECY-93-127, dated July 13, 1993 (ML003760936)

14. NUREG/CR-6451 "A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants," dated August 31, 1997 (ML082260098)

15. Letter from NEDA (R. Anderson) to USN RC, "NextEra Energy Duane Arnold, LLC Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," NG-14-0092, dated March 28, 2014(ML14092A331)

16. Letter from USNRC to NEDA (T. Vehec), "Duane Arnold Energy Center- Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (TAC No. MF3783),"

dated December 15, 2015 (ML15324A176)

17. Letter from USNRC to Listed Power Reactor Licensees, "Staff Review of High Frequency Confirmation Associated with Reevaluated Seismic Hazard in Response to March 12, 2012 50.54(f) Request for Information," dated February 18, 2016, (ML15364A544)