Text

Response to Request for Additional Information Relating to Request for Exemption from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E
	ML20282A595
Person / Time
Site:	Duane Arnold
Issue date:	10/07/2020
From:	Dean Curtland; NextEra Energy Duane Arnold
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NG-20-0069
	Download: ML20282A595 (67)
	v • d • e

NEXTeraM ENERGY~

DUANE ARNOLD October 7, 2020 NG-20-0069 10 CFR 50.12 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

Subject:

Response to Request for Additional Information Relating to Request for Exemption from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E

References:

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

2. FINAL RAI - Duane Arnold - Exemption from portions of 10 CFR 50.47 and Appendix E (EPID: L-2020-LLE-0023), dated August 5, 2020 (ML20218A772)

NextEra Energy Duane Arnold, LLC (NEDA) submitted an Exemption Request for the Duane Arnold Energy Center (DAEC) pursuant to 10 CFR 50.12 (Reference 1).

Subsequently, the NRC Staff requested additional information regarding that application (Reference 2). The enclosure to this letter contains the requested information .

Attachment 1 of the enclosure replaces, in its entirety, Attachment 1 of the Exemption Request.

This additional information does not impact the 10 CFR 50.92 evaluation of "No Significant Hazards Consideration" previously provided in the referenced application.

This letter makes no changes to existing commitments and makes no new commitments.

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

NextEra Energy Duane Arnold , LLC , 3277 DAEC Road, Palo , IA 52324

Document Control Desk NG-20-0069 Page 2 of 2 Dean Curtland Site Director, Duane Arnold Energy Center NextEra Energy Duane Arnold, LLC Enclosure cc: Regional Administrator, USNRC, Region Ill Resident Inspector, USNRC, Duane Arnold Energy Center Project Manager, USNRC, Duane Arnold Energy Center State of Iowa

ENCLOSURE TO NG-20-0069 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER Response to Request for Additional Information Relating to Request for Exemption from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E 4 pages follow

NG-20-0069 Enclosure RAl-NSIR-01 Applicable Regulation and Guidance 10 CFR 50.155, "Mitigation of beyond-design-basis events," indicates that licensees shall develop, implement, and maintain extensive damage mitigation guidelines. Specifically, 10 CFR 50.155(b )(2) states that licenses will maintain:

Strategies and guidelines to maintain or restore core cooling, containment, and spent fuel pool cooling capabilities under the circumstances associated with loss of large areas of the plant impacted by the event, due to explosions or fire, to include strategies and guidelines in the following areas:

(i) Firefighting; (ii) Operations to mitigate fuel damage; and (iii) Actions to minimize radiological release.

Licensees shall comply with the requirements of 10 CFR 50.155 until "all irradiated fuel has been permanently removed from the spent fuel pool(s)."

Issue Item #1 in Table 1, "Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)," of to the April 2, 2020 application letter states, in part, These mitigative strategies, addressing events involving a loss of SFP [spent fuel pool] cooling and/or water inventory, include implementation of SFP inventory makeup strategies required under 10 CFR 50.54(hh)(2), which will continue to be maintained to satisfy applicable License Conditions of the Renewed Facility License. (emphasis added)

Please clarify and/or correct the use of 10 CFR 50.54(hh)(2) as it relates to maintaining mitigative strategies.

DAEC Response "10 CFR 50.54(hh)(2)" has been replaced with "10 CFR 50.155" in Item #1 of Table 1. Verbiage will be added to clarify that the requirements of 10 CFR 50.155 will be met, as applicable.

Reference to 10 CFR 50.54(hh)(2) has also been updated to 10 CFR 50.155 in Item #6 of Table 1, Item #7 of Table 3 and Item #2 of Table 4.

RAl-NSIR-02 Applicable Regulation and Guidance Section IV.F.2.j in Appendix E to 10 CFR Part 50 provides the requirements for conducting emergency preparedness exercises.

1

NG-20-0069 Enclosure Issue The regulation wording for item #98 in Table 2, "Exemptions Requested from 10 CFR 50, Appendix E," of Attachment 1 to the April 2, 2020 application does not appear to be consistent with the current regulation language of Section IV.F 2.j in Appendix E to 10 CFR Part 50, specifically with the application's reference to 10 CFR 50.54(hh)(2).

Please clarify and/or correct this inconsistency.

DAEC Response The regulation wording of Item #98 in Table 2 has been updated to accurately reflect the current wording in 10 CFR 50, Appendix E, Section IV.F.2.j. This included replacing "10 CFR 50.54(hh)(2)" with "10 CFR 50.155."

RAl-NSIR-03 Applicable Regulation and Guidance Section 2.3.1, "SFP [Spent Fuel Pool] Make Capability," of Nuclear Energy Institute (NEI) guidance document NEI 06-12, "B.5.b Phase 2 & 3 Submittal Guideline," (ADAMS Accession No. ML070090060), states, in part:

Objective: Establish a flexible means of SFP makeup of at least 500 gpm

[gallons per minute] using a portable, power independent pumping capability.

Performance Attributes Item 11: The system should be capable of being deployed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from the time plant personnel diagnose that external SFP makeup is required.

Issue Item #2 in Table 5, "Staff Decommissioning Assumptions (SDAs) Comparison," to the April 2, 2020 application states, in part:

Specifically, DAEC's standard portable fire pumps deliver adequate head and flow to provide the minimum require makeup to the SFP. The evaluation indicates this equipment can be installed in approximately 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to deliver SFP makeup. The required equipment and installation procedures are required to be maintained per DAEC Operating License section 2.C.(9), "Mitigation Strategy License Condition," item (b)(7).

The capability of being deployed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from the time plant personnel diagnose that external SFP makeup is required has served as the basis for recent NRC approvals for exemptions to EP requirements for decommissioning nuclear power reactor facilities.

Please provide additional details regarding the mitigation strategies, to include the identification of the trained on-shift personnel designated for carrying out the necessary tasks and the timeframe for implementation of these mitigation strategies per NEI 06-12.

2

NG-20-0069 Enclosure DAEC Response The DAEC has demonstrated the capability to deploy the required equipment for Spent Fuel Pool make-up within the 2-hour response outlined in NEI 06-12. The validation of this capability was completed using the minimum staffing required by Technical Specifications and the same procedure and equipment (including portable diesel fire pump) currently in place for spent fuel pool make-up.

RAl-NSIR-04 Applicable Regulation and Guidance EA-12-049, "Order Modifying Licenses With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," (ADAMS Accession No. ML12056A045), requires provisions for mitigation strategies for beyond-design-basis external events.

EA-12-051, "Order Modifying Licenses With Regard to Reliable Spent Fuel Pool Instrumentation," (ADAMS Accession No. ML12056A044), requires provisions for reliable spent fuel pool indications.

Issue Item #5 in Table 4, "Industry Decommissioning Commitments (IDCs) Comparison," to the April 2, 2020 application states, in part:

SFP Level instrumentation provides indication and alarm to Control Room. This consists of two pool level instruments installed in accordance with NRC Order EA-12-049 and one pool level instrument on the main control board. (emphasis added)

Please clarify and/or correct the reference to Order EA-12-049 instead of Order EA 051.

DAEC Response Reference to NRC Order EA-12-049 has been corrected to EA-12-051 in Item #5 of Table 4.

ADDITIONAL INFORMATION NEDA had previously planned to permanently cease power operation of DAEC on October 30, 2020. DAEC was shut down on August 10, 2020 in response to a derecho storm and the decision has been made to remain permanently shut down. NEDA submitted a certification of permanent cessation of power operations to the NRC on August 27, 2020 (ML20240A067).

Calculation CAL-M 19-001, "Adiabatic Heatup Analysis of Drained Spent Fuel Pool (Zirconium Fire)", which was included in the exemption request (ML20101M779), demonstrated that the 10-hour fuel temperature rise to 900° C would be precluded by a 9-month decay time assuming the shutdown would occur on 10/31/20. The DAEC reviewed the methodology used in the calculation and determined that in order to assure adequate conservatisms existed, a 10-month 3

NG-20-0069 Enclosure period of decay would be used as the point at which the potential for the clad to reach 900 degrees C within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months following a loss of coolant in the fuel pool would not occur. DAEC reviewed CAL-M 19-001 for impact due to the earlier shutdown date, and found the earlier shutdown added additional conservatisms to the original results of the analysis. Therefore, DAEC has determined that there are no changes needed to the requested 10-month fuel decay period. Given the permanent shutdown date of August 10, 2020 and the proposed fuel decay time of 1O months, the period in which the spent fuel could heat up to clad ignition temperature within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months under adiabatic conditions would end on June 10, 2021.

In order to assure the limitations placed on the aluminum PaR spent fuel pool racks would not adversely affect the ability to mitigate the loss of spent fuel pool coolant inventory by addition of water within the 10-hour period following a postulated loss of spent fuel pool coolant, the DAEC will only use the Holtec spent fuel pool racks for the final core discharge to the spent fuel pool.

The 368 freshly discharged fuel bundles will be placed in the appropriate checkerboard pattern within the nine (9) Holtec Spent Fuel Pool Racks.

4

NG-20-0069 Attachment 1 to Enclosure Duane Arnold Energy Center Request for Exemptions from Portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and 10 CFR Part 50, Appendix E 59 pages follow

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 1 of 59

1. Summary Description Pursuant to 10 CFR 50.12 "Specific exemptions, " NextEra Energy Duane Arnold, LLC (NEDA) requests the following regulatory exemptions for the Duane Arnold Energy Center (DAEC):

Certain standards in 10 CFR 50.47(b) regarding onsite and offsite emergency response plans for nuclear power reactors;

Certain requirements of 10 CFR 50.47(c)(2) to establish Plume Exposure and Ingestion Pathway Emergency Planning Zones (EPZs) for nuclear power plants; and

Certain requirements of 10 CFR 50, Appendix E, which establishes the elements that make up the content of emergency plans.

2. 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 2), 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).

3. Detailed Description The current 10 CFR Part 50 regulatory requirements for emergency planning (developed for operating reactors) ensure the health and safety of the public while DAEC is licensed to operate. However, once the station is permanently shut down and defueled, some of these requirements exceed what is necessary to protect the health and safety of the public. In order to allow a reduction in emergency planning requirements commensurate with the hazards

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 2 of 59 associated with DAE C's permanently defueled condition, exemptions from portions of 10 CFR 50.47(b), 50.47(c)(2), and 10 CFR 50, Appendix E, are needed.

NEDA has performed an analysis indicating that, within 10 months after shutdown, the spent fuel in the spent fuel pool (SFP) will have decayed to the extent that the requested exemptions can be implemented at DAEC without any compensatory actions. This analysis is included in . Separate from this request, NEDA plans to submit a Permanently Defueled Emergency Plan, including a Permanently Defueled Emergency Action Level scheme, for NRC review and approval pursuant to 10 CFR 50.54(q)(4) and 10 CFR 50, Appendix E, Section IV.B.2. The proposed emergency plan will be based on the exemptions requested herein.

Based on the analysis detailed in Section 4 below, NEDA has concluded that the portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and 10 CFR 50, Appendix E, identified in Tables 1 and 2, will not be necessary to protect the health and safety of the public once DAEC is in the permanently defueled condition, and would be unduly burdensome. Approval of the exemptions requested in Tables 1 and 2 would not present an undue risk to the public or prevent appropriate response in the event of an emergency at DAEC.

The requested exemptions and justification for each are based on, and consistent with, Interim Staff Guidance (ISG) NSIR/DPR-ISG-02, "Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants" (Reference 1).

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 3 of 59 EXEMPTIONS TO EMERGENCY PLAN REQUIREMENTS DEFINED BY 10 CFR 50.47 AND APPENDIX E TO PART 50 NEDA requests exemptions from portions of 10 CFR 50.47(b) and (c)(2) and Appendix E to 10 CFR Part 50 to the extent that these regulations apply to specific provisions of onsite and offsite emergency planning that will no longer be applicable to DAEC once the certifications required by 10 CFR 50.82(a)(1 )(i) and (ii) have been submitted and sufficient decay of the spent fuel has occurred. The specific portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E from which exemptions are being requested are identified using strikethrough text in Table 1 (Exemptions Requested from 10 CFR 50.47(b) and (c)(2)) and Table 2 (Exemptions Requested from 10 CFR Part 50, Appendix E) below. The portions of regulation that are not identified using strikethrough text (i.e., those portions for which exemption is not being requested), will remain applicable to DAEC. Details related to specific exemption requests are provided in the Basis for Exemption column.

Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)

Item# Regulation in 10 CFR 50.47 Basis for Exemption 1 10 CFR 50.47(b): The onsite and, except as provided in In the Statement of Considerations for the Final rule - 10 CFR Part 72, paragraph (d) of this section, offsite emergency response "Emergency Planning Licensing Requirements for Independent Spent plans for nuclear power reactors must meet the following Fuel Storage Facilities (ISFSI) and Monitored Retrievable Storage standards: Facilities (MRS)," (Reference 3), the Commission responded to comments concerning offsite emergency planning for ISFSls or an MRS and concluded that, "the offsite consequences of potential accidents at an ISFSI or a MRS installation would not warrant establishing Emergency Planning Zones."

In a nuclear power reactor's permanently defueled state, the accident risks are more similar to an ISFSI or MRS than an operating nuclear power plant. The draft proposed rulemaking in SECY-00-0145 (Reference 5) suggested that after at least one year of spent fuel decay time, the decommissioning licensee would be able to reduce its EP program to one similar to that required for an MRS under 10 CFR 72.32(a) and additional EP reductions would occur when: (1) approximately five years of spent fuel decay time has elapsed; or (2) a licensee has demonstrated that the decay heat level of spent fuel in the pool is low enough that the fuel would not be susceptible to a zirconium fire for all spent fuel configurations. The EP program would be similar to that required for an ISFSI under 10 CFR 72.32(a) when fuel stored in the SFP has more than five years of decay time and would not change substantially when all the fuel is transferred from the SFP to an onsite

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 4 of 59 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)

Item# Regulation in 10 CFR 50.47 Basis for Exemption ISFSI. Exemptions from offsite EP requirements have been approved when the specific site analyses show that at least ten hours is available from a partial drain down event where cooling of the spent fuel is not effective until the hottest fuel assembly reaches 900°C. Because ten hours allows sufficient time to initiate mitigative actions to prevent a zirconium fire in the SFP or to initiate ad hoc offsite protective actions, offsite EP plans are not necessary for these permanently defueled nuclear power plant licensees.

NEDA has performed an analysis (Attachment 1) indicating that after the spent fuel has decayed for 10 months, for beyond design basis events where the SFP is fully drained and air cooling is not credited, 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is available to take mitigative actions or, if needed, implement offsite protective actions using a comprehensive approach to emergency planning from the time spent fuel cooling is lost until the hottest fuel assembly reaches a temperature of 900°C.

Additional NEDA analysis (Attachment 2) has demonstrated that within 9 months after permanent cessation of power operations, the radiological consequences of the postulated accident will not exceed the limits of the U.S. Environmental Protection Agency's (EPA's)

Protective Action Guides (PAGs) at the Exclusion Area Boundary (EAB). (Reference 7)

NEDA maintains procedures and strategies for the movement of any necessary portable equipment that will be relied upon for mitigating -

the loss of SFP water. These mitigative strategies, addressing events involving a loss of SFP cooling and/or water inventory, include implementation of SFP inventory makeup strategies required under 10 CFR 50.155 for a permanently shut down and defueled facility, which will continue to be maintained to satisfy

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 5 of 59 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)

Item# Regulation in 10 CFR 50.47 Basis for Exemption applicable License Conditions of the Renewed Facility License. These diverse strategies provide defense-in-depth and ample time to provide makeup water or spray to the SFP prior to the onset of zirconium cladding ignition when considering very low probability beyond design basis events affecting the SFP. The on-shift individuals described in the Permanently Defueled Emergency Plan will be able to implement the necessary tasks within the required timeframe.

2 10 CFR 50.47(b)(1): Primary responsibilities for emergency Refer to basis for 10 CFR 50.47(b).

response by the nuclear facility licensee and by State and local organizations 'Nithin the Emergency Planning Zones have been assigned, the emergency responsibilities of the various supporting organizations have been specifically established, and each principal response organization has staff to respond and to augment its initial response on a continuous basis.

3 10 CFR 50.47(b)(2): On-shift facility licensee responsibilities No exemption is requested.

for emergency response are unambiguously defined, adequate staffing to provide initial facility accident response in key functional areas is maintained at all times, timely augmentation of response capabilities is available and the interfaces among various onsite response activities and offsite suooort and response activities are specified.

4 10 CFR 50.47(b)(3): Arrangements for requesting and Discontinuing offsite emergency planning activities and reducing the effectively using assistance resources have been made, scope of onsite emergency planning is acceptable given the aFFangements to accommoElate State anEI local staff at the significantly reduced offsite consequences once DAEC is in the licensee's EmeFgency Opemtions Facility have been maEle, permanently defueled condition. The DAEC Emergency Plan will and other organizations capable of augmenting the planned continue to maintain arrangements for requesting and using assistance response have been identified. resources from offsite support organizations.

Decommissioning power reactors present a low likelihood of any credible accident resulting in radiological releases requiring offsite protective measures because of the permanently shut down and

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 6 of 59 Table 1 Exemptions Requested from 10 CFR 50.47(b} and 50.47(c)(2}

Item# Regulation in 10 CFR 50.47 Basis for Exemption defueled status of the reactor. An Emergency Operations Facility (EOF) is not required. The control room or other location can provide for the communication and coordination with offsite organizations for the level of support required.

Offsite emergency measures are limited to support provided by local police, fire departments, and ambulance and hospital services as appropriate.

Also refer to basis for 10 CFR 50.47(b).

5 10 CFR 50.47(b)(4): A standard emergency classification and NEDA will adopt the Permanently Defueled Emergency Action Levels action level scheme, the basis of which include facility (EALs) detailed in Appendix C of Nuclear Energy Institute (NEI) 99-01, system and effluent parameters, is in use by the nuclear "Development of Emergency Action Levels for Non-Passive Reactors,"

facility licensee, aAEI State aAEI lesal FSSfleAse fllaAs sall feF Revision 6 (Reference 4), endorsed by the NRC in a letter dated March FeliaA68 9R iRfeFmatieR flF9ViEleEI ey fasility liseRsees feF 28, 2013 (Reference 6). No offsite protective actions are anticipated to EleteFmiRatieAs ef miAimum iRitial effsite FeSfleAse measurns. be necessary, so classification above the Alert level will no longer be required.

Also refer to basis for 10 CFR 50 .47 (b).

6 10 CFR 50.47(b)(5): Procedures have been established for Per SECY-00-0145 (Reference 5), after approximately 1 year of spent notification, by the licensee, of State and local response fuel decay time, the NRC staff believes an exception to the offsite EPA organizations and for notification of emergency personnel by PAG standard is justified for a zirconium fire scenario considering the all organizations; the content of initial and followup messages low likelihood of this event together with time available to take to response organizations aREI tt:ie f>Uelis has been mitigative or protective actions between the initiating event and before established; aREI meaRs te FJFeviEle eaFly AetifisatieR aREI sleaF the onset of a postulated fire. The spent fuel scoping study (Reference iAstFustieA te tt:ie fl9flUlase witRiA tt:ie fllume 8*fl9SUFe 22) provides that depending on the size of the pool liner leak, releases f)att:ivvay EmeFgeAsy PlaRAiAg ZeRe t:iave eeeR estaelist:ieEI. could start anywhere from eight hours to several days after the leak starts, assuming that mitigation measures are unsuccessful. If 10 CFR 50.155 type of mitigation measures are successful, releases could only occur during the first several days after the fuel came out of the reactor.

NEDA analysis shows that after the spent fuel has decayed for 10 months, if the SFP is fully drained and air coolinq is not credited, 10

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 7 of 59 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)

Item# Regulation in 10 CFR 50.47 Basis for Exemption hours are available until the hottest fuel assembly reaches a temperature of 900°C. This is adequate time to take mitigative or, if needed, offsite protective actions using a comprehensive approach to emergency planning. No offsite protective actions are anticipated to be necessary. Therefore, offsite EP plans are not necessary for permanently defueled nuclear power plants.

Also refer to basis for 10 CFR 50.47(b).

7 10 CFR 50.47(b)(6): Provisions exist for prompt Refer to basis for 10 CFR 50.47(b).

communications among principal response organizations to emerQencv personnel =~~ :~ t!;:::: ;: *"'!:::::.

8 10 CFR 50.47(b)(7): IAfoFmatieA is maEle availaele te tF!e Refer to basis for 10 CFR 50.47(b).

131::1l31ie eA a 13eFieElie 13asis eA F!e1N tF!ey will 13e AetifieEI aAEI 1NF!at tF!eiF iAitial aetieAs sF!e1::1IEI ee iA aA eFReF§leAey (e.§1.,

listeAiA§! te a leeal bmaEleast statieA aAEI reFRaiAiA§! iAEleeFS),

[T]he principal points of contact with the news media for dissemination of information during an emergency (iAel1::1EliA§!

tF!e 13hysieal leeatieAs) are established in advance, and procedures for coordinated dissemination of information to the public are established.

9 10 CFR 50.47(b)(8): Adequate emergency facilities and No exemption is requested.

equipment to support the emergency response are provided and maintained.

10 10 CFR 50.47(b)(9): Adequate methods, systems, and Refer to basis for 10 CFR 50.47(b).

equipment for assessing and monitoring actual or potential effs+te consequences of a radiological emergency condition are in use.

11 10 CFR 50.47(b)(10): A range of protective actions has been In the unlikely event of a SFP accident, the iodine isotopes which developed for tF!e 13l1::1me ex13es1::1re 13athway EPZ foF contribute to an off-site dose from an operating reactor accident are not emergency workers and the public. IA Elevele13iA§! tF!is F8A§le present, so potassium iodide (Kl) distribution off-site would no longer ef aetieAs, ceAsiElemtieA has beeA §!iveA te evac1::1atieA, serve as an effective or necessary supplemental protective action.

- *- - ***~, - *-, - -~

~'--*~--=-- ~-..J - - ~ ~ *-""1 .... ---~ ~- ~ .... *- **-

- - . Protective actions will be maintained for emergency workers and any

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 8 of 59 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)

Item# Regulation in 10 CFR 50.47 Basis for Exemption 1::1se ef 13etassi1::1FR ieeliele ~KJ~, as a1313Fe13Fiate. ElJae1::1atieR offsite emergency responders who would respond to the site.

tiFRe estiFRates l=lave 13eeR elevele13eel 13y a1313lieaRts aREI lieeRsees. bieeRsees sl=lall 1::113elate tl=le evae1::1atieR tiFRe The Commission responded to comments in its Statement of estiFRates SR a 13eFieelie 13asis. G1::1ieleliRes feF tl=le el=leiee ef Considerations for the Final Rule for emergency planning requirements 13Feteeth1e aetieRs El1::1FiR§ aR eFReF§eRey, eeRsisteRt witl=I for ISFSls and MRS facilities (60 FR 32435) (Reference 3), and P:eeleFal §t1ielaRee, aFe elevele13eel aREI iR 13laee, aREI 13Feteeti1.ie concluded that, "the offsite consequences of potential accidents at an aetieRs foF tl=le iR§estieR e*13es1::1Fe 13att:i11.1ay EP;6 a1313Fe13Fiate ISFSI or a MRS would not warrant establishing Emergency Planning te tl=le leeale l=la 1*1e 13eeR elevele13eel. Zones." Additionally, in the Statement of Considerations for the Final Rule for EP requirements for ISFSI and for MRS facilities (60 FR 32430) (Reference 3), the Commission responded to comments concerning site-specific emergency planning that includes evacuation of surrounding population for an ISFSI, not at a reactor site, and concluded that, "The Commission does not agree that as a general matter emergency plans for an ISFSI must include evacuation planning."

Because the NRC concludes that evacuation planning is not needed for a decommissioning reactor site that meets the criteria for an exemption from offsite EP requirements as discussed in the exemption from 10 CFR 50.47(b), evacuation time estimates are also not needed.

Also refer to the basis for 10 CFR 50.47(b) detailing the low likelihood of any credible accident resulting in radiological releases requiring offsite protective measures and the basis for Section IV.1 exemptions for a discussion on the similarity between a permanently defueled reactor and a non-power reactor.

12 10 CFR 50.47(b)(11 ): Means for controlling radiological No exemption is requested.

exposures, in an emergency, are established for emergency workers. The means for controlling radiological exposures shall include exposure guidelines consistent with EPA Emergency Worker and Lifesaving Activity Protective Action Guides.

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 9 of 59 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2)

Item# Regulation in 10 CFR 50.47 Basis for Exemption 13 10 CFR 50.47(b)(12): Arrangements are made for medical No ex~mption is requested.

services for contaminated injured individuals.

14 10 CFR 50.47(b)(13): General plans for recovery and reentry No exemption is requested.

are developed.

15 10 CFR 50.47(b)(14): Periodic exercises are (will be) No exemption is requested.

conducted to evaluate major portions of emergency response capabilities, periodic drills are (will be) conducted to develop and maintain key skills, and deficiencies identified as a result of exercises or drills are (will be) corrected.

16 10 CFR 50.47(b)(15): Radiological emergency response No exemption is requested.

training is provided to those who may be called on to assist in an emeraencv.

17 10 CFR 50.47(b)(16): Responsibilities for plan development No exemption is requested.

and review and for distribution of emergency plans are established, and planners are properly trained.

18 10 CFR 50.47(c)(2): GeReFally, U:ie ph.1R=ie e*pes1:1Fe patl=lway Refer to basis for 10 CFR 50.47(b)(10).

eP2; feF R1:1eleaF pe11.ieF plaRts sl=lall eeRsist ef aR aFea aee1:1t

~ Q R=iiles ~~ e kffi~ iR FaElil:IS aREI tl=le iR§estieR patl=w.iay eP2; sl=lall eeRsist ef aR aFea aee1:1t §Q R=iiles ~gg l'<ffi~ iR FaEli1:1s.

+Fie e*aet siZ!:e aREI eeRfi§1:1FatieR ef tl=le eP2;s s1:1FFe1:1REliR§ a paFtie1:1laF Rl:leleaF pe11.1eF FeaeteF SRall ee EleteFffiiReEI iR FelatieR te leeal eR:ieF§eRey FespeRse ReeEls aREI eapasilities as tl=ley aFe affeeteEI lay s1:1el=l eeRElitieRs as Eleffie§Fapl=ly, tepe§mpl=ly, laREI cl=lameteFisties, aeeess m1:1tes, aREI j1:1FisElietieRal ee1:1RElaFies. The size of the EPZs aJ.se-may be determined on a case-by-case basis for gas cooled nuclear reactors and for reactors with an authorized power level less than 250 MW thermal. +Fie plaRs feF tl=le iR§estieR patR1Nay sl=lall fec1:1s eR s1:1cl=l aetieRs as aFe appmwiate te pmteet tl=le

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Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 10 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption 19 10 CFR 50 App E I No exemption is requested.

Ill. The Final Safety Analysis Report; Site Safety Analysis Report The final safety analysis report or the site safety analysis report for an early site permit that includes complete and integrated emergency plans under§ 52.17(b)(2)(ii) of this chapter shall contain the plans for coping with emergencies.

The plans shall be an expression of the overall concept of operation; they shall describe the essential elements of advance planning that have been considered and the provisions that have been made to cope with emergency situations. The plans shall incorporate information about the emergency response roles of supporting organizations and offsite agencies. That information shall be sufficient to provide assurance of coordination among the supporting groups and with the licensee. The site safety analysis report for an early site permit which proposes major features must address the relevant provisions of 10 CFR 50.47 and 10 CFR part 50, appendix E, within the scope of emergency preparedness matters addressed in the major features. The plans submitted must include a description of the elements set out in Section IV for the emergency planning zones (EPZs) to an extent sufficient to demonstrate that the plans provide reasonable assurance that adequate protective measures can and will be taken in the event of an emerqencv.

20 10 CFR 50 App E Following docketing of its "Certification of Permanent Removal of Fuel from the Reactor Vessel," in accordance with 10 CFR 50.82(a)(1 )(ii),

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 11 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption IV Content of Emergency Plans DAEC will become a permanently shut down facility with spent fuel stored in the SFP. In the EP Final Rule (76 FR 72596, Nov. 23, 2011)

1. The applicant's emergency plans shall contain, but not (Reference 8), the NRC defined "hostile action" as, in part, an act necessarily be limited to, information needed to demonstrate directed toward a nuclear power plant or its personnel. This definition is compliance with the elements set forth below, i.e., based on the definition of "hostile action" provided in NRC Bulletin organization for coping with radiological emergencies, 2005-02. NRC Bulletin 2005-02 was not applicable to nuclear power assessment actions, activation of emergency organization, reactors that have permanently ceased operations and have certified notification procedures, emergency facilities and equipment, that fuel has been removed from the reactor vessel. The NRC excluded training, maintaining emergency preparedness, and recovery, non-power reactors (NPRs) from the definition of "hostile action" at that and onsite protective actions during hostile action. In time because an NPR is not a nuclear power plant and a regulatory addition, the emergency response plans submitted by an basis had not been developed to support the inclusion of NPR in that applicant for a nuclear power reactor operating license under definition. Similarly, a decommissioning power reactor or ISFSI is not a this part, or for an early site permit (as applicable) or "nuclear reactor" as defined in the NRC's regulations.

combined license under 10 CFR part 52, shall contain information needed to demonstrate compliance with the The following similarities between DAEC and NPRs show that the standards described in § 50.47(b), and they will be evaluated DAEC facility should be treated in a similar fashion as an NPR. Similar against those standards. to NPRs, DAEC will pose lower radiological risks to the public from accidents than do power reactors because: (1) DAEC will be a permanently shut down facility (with fuel stored in the SFP and ISFSI) and will no longer generate fission products; 2) Fuel stored in the DAEC SFP will have lower decay heat resulting in lower risk of fission product release in the event of a non-credible boil off or drain down event; and 3) no credible accident at DAEC will result in radiological releases requirinQ offsite protective actions.

21 IV.2 +l=lis nuclear po1.ver reactor license applicant sl=lall also Refer to basis for 10 CFR 50.47(b)(10).

provide an analysis of tl=le tirne reEluired to evacuate 1.tarious sectors and distances v.iitl=lin tl=le plurne e*posure pathway 6P~ for transient and perrnanent populations, using tl=le most recent U.S. Census Bureau data as of tl=le date tl=le applicant ro.

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22 IV. 3 Nuclear pov.ier reactor licensees sl=lall use ~H~C Refer to basis for 10 CFR 50.47(b)(10).

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Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 12 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E I Basis for Exemption 23 I IV.4 VVithin 365 davs of the later of the date of the availabilitv I Refer to basis for 10 CFR 50.47(b)(10).

of the most recent decennial census data from the U.S.

Census Bureau or December 23, 2011, nuclear pmver reactor licensees shall develop an ETE analysis using the decennial data and submit it under§ 50.4 to the NRG. These licensees shall submit this ETE analysis to the NRG at least 180 days before using it to form protective action recommendations and providing it to State and local governmental authorities for use in developing offsite 24 I JV.5 During the years between decennial censuses, nuclear I Refer to basis for 10 CFR 50.47(b)(10).

pm.ver reactor licensees shall estimate EPZ permanent resident population changes once a year, but no later than 365 days from the date of the previous estimate, using the most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. These licensees shall maintain these estimates so 25 I IV.6 If at any time during the decennial period, the EPZ I Refer to basis for 10 CFR 50.47(b)(10).

permanent resident population increases such that it causes the longest ETE value for the 2 mile zone or 5 mile zone, including all affected Emergency Response Planning Areas, or for the entire 10 mile EPZ to increase by 25 percent or 30 minutes, \Nhichever is less, from the nuclear power reactor licensee's currently NRG approved or updated ETE, the

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 13 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption tl=lat 13e131:1latieR iRcFease. +!=le liceRsee sl=lall s1:1l3ffiit tl=le 1:113ElateEI E+E aRalysis te tl=le NRG 1:1REleF § §Q.4 Re lateF tl=laR

3@§ Elays afteF U:ie liceRsee's EleteFffiiRatieR tl=iat tl=ie cFiteFia feF 1:113ElatiR§ tl=le E+E !=lave eeeR ffiet aREI at least ~ gg Elays eefeFe 1:1siR§ it te feFffi 13Feteetive actieR FeceffiffieRElatieRs aREI 13rnviEliR§ it te State aREI lecal §9VeFRffieRtal a1:1tl=ieFities

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26 IV 7 After an applicant for a combined license under part 52 No exemption is requested. NEDA is not an applicant for a combined of this chapter receives its license, the licensee shall conduct license, and therefore, this regulation is not applicable to DAEC.

at least one review of any changes in the population of its EPZ at least 365 days prior to its scheduled fuel load. The licensee shall estimate EPZ permanent resident population changes using the most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. If the EPZ permanent resident population increases such that it causes the longest ETE value for the 2-mile zone or 5-mile zone, including all affected Emergency Response Planning Areas, or for the entire 10-mile EPZ, to increase by 25 percent or 30 minutes, whichever is less, from the licensee's currently approved ETE, the licensee shall update the ETE analysis to reflect the impact of that population increase. The licensee shall submit the updated ETE analysis to the NRC for review under§ 50.4 of this chapter no later than 365 days before the licensee's scheduled fuel load.

27 A. Organization No exemption is requested.

The organization for coping with radiological emergencies shall be described, including definition of authorities, responsibilities, and duties of individuals assigned to the licensee's emen::iencv on::ianization and the means for

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 14 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption notification of such individuals in the event of an emergency.

Specifically, the following shall be included:

28 A.1. A description of the normal plant operating organization. Following docketing of the certifications required by 10 CFR 50.82(a)(1), DAEC will not be a facility that can be operated to generate electrical power. Therefore, DAEC will not have a "plant operating organization." Rather, the station will be maintained by a defueled on-shift staff.

29 A.2. A description of the onsite emergency response No exemption is requested.

organization (ERO) with a detailed discussion of:

a. Authorities, responsibilities, and duties of the individual(s) who will take charge during an emergency;

b. Plant staff emergency assignments;

c. Authorities, responsibilities, and duties of an onsite emergency coordinator who shall be in charge of the exchange of information with offsite authorities responsible for coordinating and implementing offsite emergency measures.

30 A.3. A EleseFiption, 13y position anEI f1::1netion to 13e peFfoFFAeEI, The number of staff at DAEC during the decommissioning process will of tl=le lieensee's l=leaE1ei1::1aFteFs peFsonnel v.il=lo will 13e sent to be small but commensurate with the need to safely store spent fuel at tl=le plant site to a1::1gFAent tl=le onsite emeFgeney OFganization. the facility in a manner that is protective of public health and safety.

Decommissioning sites typically have a level of emergency response that does not require response by headquarters personnel.

31 A.4. Identification, by position and function to be performed, Analyses have been developed indicting that, within 10 months after of persons within the licensee organization who will be shutdown, no credible accident at DAEC will result in radiological responsible for making e#site dose projections and a releases requiring offsite protective actions.

description of how these projections will be made and the results transmitted to State and local authorities, NRC, and NEDA will maintain the capability to determine if a radiological release other appropriate aovernmental entities. is occurrina. If a release is occurring, then NEDA will promptly

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 15 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption communicate that information to offsite authorities for their consideration. The offsite organizations are responsible for deciding what, if anv, protective actions should be taken.

32 A 5. IEleAtifieatieA, ey 13esitieA aAEI f1:metieA ta ee 13eFfeFA=ieEI, The time available to initiate compensatory actions in the event of a ef eU:ieF eFR13leyees ef tl=ie !ieeAsee witl=i s13eeia! q1;;1alifieatieAs loss of SFP cooling or inventory precludes the need to identify and feF ee13iA§ witl=i eA=ieF§eAey eeAElitieAs tl=iat FRay aFise. Gtl=ieF describe the special qualification of these individuals in the emergency 13eFS9AS witl=i s13eeial q1::1alifieatieAS, Sl::IGA as G9ASl::lltaAts, 1NA9 plan. The number of staff at DAEC, once it is in the permanently aFe Aet eFR13leyees e:f tl=ie lieeAsee aAEI 1.Yl=ie A=iay ee ealleEI defueled state, will be small but commensurate with the need to 1::113eA :feF assistaAee :feF eFReF§eAeies sl=iall alse ee iEleAtifieEI. operate the facility in a manner that is protective of public health and Tl=ie s13eeial q1;;1alifieatieAs ef tl=iese 13eFSens sl=ial! ee safety.

EleseFi13eEI.

33 A 6. A description of the local offsite services to be provided No exemption is requested.

in support of the licensee's emergencv oroanization.

34 A.7. By J1::1Ae 23, 2014, identification of, aAEI a EleseFi13tieA e:f A decommissioning power reactor has a low likelihood of a credible tJ::t.e-assistance expected from, appropriate State, local and accident resulting in radiological releases requiring offsite protective Federal agencies with responsibilities for coping with measures. For this reason and those described in the basis for emergencies, including l=iestile aetieA at tl=ie site. FeF Section IV.1 of 10 CFR Part 50, Appendix E, a decommissioning 131::1F13eses e:f tl=iis a1313eAElix, "l=iestile aetieA" is ElefiAeEI as an power reactor is not a facility that falls within the definition of "hostile act directed toward a nuclear power plant or its personnel action."

that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an Similarly, for security, risk insights can be used to determine which end. This includes attack by air, land, or water using guns, targets are important to protect against sabotage. A level of security explosives, projectiles, vehicles, or other devices used to commensurate with the consequences of a sabotage event is deliver destructive force. required and is evaluated on a site-specific basis.

Although, the analysis described above and in the basis for 10 CFR Part 50, Appendix E, Section IV.1 provides a justification for exempting NEDA from "hostile action" related requirements, some EP requirements for security-based events will be maintained. The classification of security-based events, notification of offsite authorities, and coordination with offsite agencies under a comprehensive emerqencv manaqement plan concept will still be

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 16 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption required.

NEDA will maintain appropriate actions for the protection of onsite personnel in a security-based event. The scope of protective actions will be appropriate for the defueled plant status, but will not be the same as actions necessary for an operating power plant. Although the NRC has previously exempted decommissioning power reactors from "hostile action" considerations, the DAEC Physical Security Plan will continue to provide high assurance against a potential security event impacting a designated target set. Therefore, some EP requirements for security-based events are maintained. Protective actions are maintained for onsite personnel through the classification of security-based events, notification of offsite authorities, and coordination of offsite response organizations (i.e., local law enforcement, firefighting, medical assistance) onsite under a comprehensive emeroencv manaoement plan.

35 A 8. leleRtifieatieR ef U1e State aRelfoF leeal e#ieials Offsite emergency measures are limited to support provided by local Fesi:;ieRsiele feF i:;ilaRRiR§ feF, eFeleFiR§, aRel eeRtFelliR§ police, fire departments, and ambulance and hospital services as ai:;ii:;iFei:;iFiate i:;iFeteetive aetieRs, iRelueliR§ evaeuatieRs wheR appropriate. Because analyses have been developed indicating that ReeessaF)'. within 10 months after shutdown, no credible accident at DAEC will result in radiological releases requiring offsite protective actions, protective actions such as evacuation should not be required.

Also refer to basis for 10 CFR 50.47(b)(10).

36 A 9. By QeeemeeF 24, 2Q~2, feF RueleaF i:;ie1.-.ieF FeaeteF Responsibilities of the on-shift and emergency response personnel lieeRsees, a eletaileel aRalysis elemeRstFatiR§ that eR shift will be detailed in the Permanently Defueled Emergency Plan and i:;ieFS9RRel assi§Reel emeF§eRey i:;ilaR imi:;ilemeRtatieR implementing procedures and will be regularly tested through drills and fuRetieRs aFe Ret assi§Reel Fesi:;ieRsieilities that weulel i:;iFeveRt exercises, audited, and inspected by NEDA and the NRC. The the timely i:;ieFfeFmaRee ef theiF assi§Reel fuRetieRs as duties of the on-shift personnel at a decommissioning reactor facility si:;ieeifieel iR the emeF§eRey i:;ilaR. are not as complicated and diverse as those for an operating power reactor.

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 17 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption In the EP Final Rule (Reference 8), the NRC acknowledged that the staffing analysis requirement was not necessary for non-power reactor licensees because staffing at non-power reactors is generally small, which is commensurate with operating the facility in a manner that is protective of the public health and safety. The minimal systems and equipment needed to maintain the spent nuclear fuel in the spent fuel pool or in a dry cask storage system in a safe condition requires minimal personnel and is governed by Technical Specifications.

Because of the slow rate of the event scenarios postulated in the design basis accident and postulated beyond design basis accident analyses, and because the duties of the on-shift personnel at a decommissioning reactor facility are not as complicated and diverse as those for an operating reactor, significant time is available to complete actions necessary to mitigate an emergency without impeding timely performance of emergency plan functions. For all of these reasons, it can be concluded that a decommissioning nuclear power plant is exempt from the requirement of 10 CFR Part 50, Appendix E, Section IV.A.9.

37 B. Assessment Actions NEDA will develop EALs consistent with the Permanently Defueled EALs detailed in Appendix C of NEI 99-01, Revision 6 (Reference 4).

8.1. The means to be used for determining the magnitude of, NEDA proposes to continue to review EALs with the State of Iowa on and for continually assessing the impact of, the release of an annual basis. However, based upon the reduced scope of EALs for radioactive materials shall be described, including the permanently defueled facility, the scope of the annual review of emergency action levels that are to be used as criteria for EALs is expected to be limited (informal mailings, etc.).

determining the need for notification and participation of local and State agencies, the Commission, and other Federal Also refer to basis for 10 CFR Part 50, Appendix E, Section IV.1.

agencies, and the emergency action levels that are to be used for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety. The emergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and offsite monitorinQ.

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 18 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption By June 20, 2012, for nuclear power reactor licensees, these action levels must include hostile action that may adversely affect the nuclear power plant. The initial emergency action levels shall be discussed and agreed on by the applicant or licensee and state and local governmental authorities, and approved by the NRC. Thereafter, emergency action levels shall be reviewed with the State and local governmental authorities on an annual basis.

38 B.2. A licensee desiring to change its entire emergency No exemption is requested.

action level scheme shall submit an application for an amendment to its license and receive NRC approval before implementing the change. Licensees shall follow the change process in§ 50.54(q) for all other emergency action level changes.

39 C. Activation of Emergency Organization The Permanently Defueled EALs, developed consistent with Appendix C of NEI 99-01, Revision 6, will be adopted, as previously described.

C.1. The entire spectrum of emergency conditions that This scheme eliminates the Site Area Emergency and General involve the alerting or activating of progressively larger Emergency event classifications. Additionally, the need to base EALs segments of the total emergency organization shall be on containment parameters is no longer appropriate. The EAL scheme described. The communication steps to be taken to alert or presented in NEI 99-01 Revision 6 was endorsed by the NRC in a letter activate emergency personnel under each class of dated March 28, 2013 (Reference 6). No offsite protective actions are emergency shall be described. Emergency action levels anticipated to be necessary, so classification above the Alert level is no (based not only on onsite and offsite radiation monitoring longer required. In the event of an accident at a defueled facility that information but also on readings from a number of sensors meets the conditions for relaxing of emergency planning requirements, that indicate a potential emergency, such as the pressure in there will be available time for event mitigation, and if necessary, containment and the response of the Emergency Core implementation of offsite protective actions using a comprehensive Cooling System) for notification of offsite agencies shall be approach to emergency planning. Refer to the basis for 10 CFR described. The existence, but not the details, of a message 50.47(b) detailing the low likelihood of any credible accident resulting in authentication scheme shall be noted for such agencies. The radiological releases requiring offsite protective measures.

emergency classes defined shall include: (1) Notification of unusual events, (2) alert, (3) site area emergency, and (4) Containment parameters will not provide an indication of the conditions aeneral emeraencv. These classes are further discussed in at a defueled facilitv and emergency core cooling systems will no

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 19 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption NUREG-0654/FEMA-REP-1. longer be required. Other indications such as SFP level or temperature will be used while there is spent fuel in the SFP.

In the Statement of Considerations for the Final Rule for EP requirements for ISFSls and for MRS facilities (60 FR 32430)

(Reference 3), the Commission responded to comments concerning a general emergency at an ISFSI and MRS, and concluded that, " ... an essential element of a General Emergency is that a release can be reasonably expected to exceed EPA Protective Action Guidelines exposure levels off site for more than the immediate site area."

The probability of a condition reaching the level above an emergency classification of Alert is very low. In the event of an accident at a defueled facility that meets the criteria for relaxation of EP requirements, there will be time available to initiate mitigative actions to protect the public.

As stated in NUREG-1738 (Reference 11 ), for instances of small SFP leaks or loss of cooling scenarios, these events evolve very slowly and generally leave many days for recovery efforts. Offsite radiation monitoring will be performed as the need arises. Due to the decreased risks associated with defueled plants, offsite radiation monitoring systems are not required.

40 C.2. By dtiRe 2G, 2G12, RtieleaF 13e1.veF FeaeteF licensees shall In the Proposed Rule (74 FR 23254) (Reference 20) to amend certain establish and maintain the capability to assess, classify, and emergency planning requirements for 10 CFR Part 50, the NRC asked declare an emergency condition withiR 15 miRtites after the for public comment on whether the NRC should add requirements for availability of indications to plant operators that an non-power reactor licensees to assess, classify, and declare an emergency action level has been exceeded and shall emergency condition within 15 minutes and promptly declare an promptly declare the emergency condition as soon as emergency condition. The NRC received several comments on these possible following identification of the appropriate emergency issues. The NRC believed there may be a need for the NRC to be classification level. Licensees shall not construe these criteria aware of security related events early on so that an assessment can be as a orace period to attempt to restore plant conditions to made to consider the likelihood that the event is part of a laroer

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 20 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # I 10 CFR Part 50, Appendix E I Basis for Exemption avoid declaring an emergency action due to an emergency coordinated attack. However, the NRC determined that further analysis action level that has been exceeded. Licensees shall not and stakeholder interactions are needed prior to changing the construe these criteria as preventing implementation of requirements for non-power reactor licensees. Therefore, the NRC did response actions deemed by the licensee to be necessary to not include requirements in the 2011 EP Final Rule (Reference 8) for protect public health and safety provided that any delay in non-power reactor licensees to assess, classify, and declare an declaration does not deny the State and local authorities the emergency condition within 15 minutes and promptly declare an opportunity to implement measures necessary to protect the emergency condition.

public health and safety.

NEDA will maintain the capability to assess, classify, and declare an emergency condition within 30 minutes after the availability of indications to operators that an EAL threshold has been reached.

Emergency declaration is required to be made as soon as conditions warranting classification are present and recognizable, but within 30 minutes in all cases of conditions being present. In the permanently defueled condition, the rapidly developing scenarios associated with events initiated during reactor power operation are no longer credible.

The consequences resulting from the only remaining events (e.g., fuel handling accident) develop over a significantly longer period. As such, the 15 minute requirement to classify and declare an emergency is unnecessarily restrictive.

Refer to basis in section IV.1 for discussion on the similarity between a permanently defueled reactor and a non-power reactor for the low likelihood of any credible accident resulting in radiological releases requirinq offsite protective measures.

41 D. Notification Procedures Refer to basis for 10 CFR 50.47(b) and 10 CFR 50.47(b)(10).

D.1. Administrative and physical means for notifying local, State, and Federal officials and agencies and agreements reached with these officials and agencies for the prompt notification of the public and for public evacuation or other

,.., ................ :e measures, should they become necessary, shall

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 21 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption be described. This description shall include identification ef.-

the appropriate officials, by title and agency, of the State and local aovernment aaencies v.rithin the EPZs 42 D.2. Provisions shall be described for yearly dissemination to Refer to basis for 10 CFR Part 50, Appendix E, Section IV.D.1.

the public within the plume exposure pathv'Jay EPZ of basic emergency planning information, such as the methods and times required for public notification and the protective actions planned if an accident occurs, general information as to the nature and effects of radiation, and a listing of local broadcast stations that *.viii be used for dissemination of 43 D.3. A licensee shall have the capability to notify responsible While the capability needs to exist for the notification of offsite State and local governmental agencies within 15 minutes government agencies within a specified time period, previous after declaring an emergency. The licensee shall exemptions have allowed for extending the State and local government demonstrate that the appropriate governmental authorities agencies' notification time up to 60 minutes based on the site-specific have the capability to mal<e a public alerting and notification justification provided.

decision promptly on being informed by the licensee of an emergency condition. Prior to initial operation greater than 5 NEDA proposes to complete emergency notification within 60 minutes percent of rated thermal po1.ver of the first reactor at a site, after an emergency declaration or a change in classification to the each nuclear power reactor licensee shall demonstrate that State of Iowa and local government agencies. This timeframe is administrative and physical means have been established for consistent with the 10 CFR 50. 72(a)(3) notification to the NRC and is alerting and providing prompt instructions to the public V'Jithin appropriate because in the permanently defueled condition, the rapidly the plume and exposure path\vay EPZ. The design objective developing scenarios associated with events initiated during reactor of the prompt public alert and notification system shall be to power operation are no longer credible and there is no need for State have the capability to essentially complete the initial alerting or local response organizations to implement any protective actions.

and initiate notification of the public within the plume The DAEC Emergency Plan includes primary and backup means for exposure pathway EPZ v.iithin about 15 minutes. The use of conducting the required notifications.

this alerting and notification capability 1.vill range from

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 22 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption minutes of the time that State and local officials are notified and responsibilities have historically involved coordination with the that a situation exists requiring urgent action) to the more State of Iowa and local counties and towns. Decommissioning-related likely events \Vhere there is substantial time available for the emergency plan submittals for DAEC have been discussed with offsite appropriate governmental authorities to make a judgment response organizations since NEDA provided notification that it would vvhether or not to activate the public alert and notification permanently cease power operations. These discussions have system. The alerting and notification capability shall addressed changes to onsite and offsite emergency preparedness additionally include administrative and physical means for a throughout the decommissioning process, including the proposed 60-backup method of public alerting and notification capable of minute notification to the State of Iowa and local government agencies.

being used in the event the primary method of alerting and Emergency management officials have not objected to the proposed notification is unavailable during an emergency to alert or notifications.

notify all or portions of the plume exposure path .vay EPZ 1

population. The backup method shall have the capability to NEDA analyses demonstrate that 10 months after permanent alert and notify the public within the plume exposure path Nay1 cessation of power operations, no remaining postulated accidents at EPZ, but does not meet the 15 minute design objective for DAEC will result in radiological releases requiring offsite protective the primary prompt public alert and notification system. actions, or in the event of beyond design basis accidents, 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is VVhen there is a decision to activate the alert and notification available to take mitigative actions, and if needed, implement offsite system, the appropriate governmental authorities will protective actions using a comprehensive emergency management determine whether to activate the entire alert and notification plan. Therefore, there is no need to maintain an Alert and Notification system simultaneously or in a graduated or staged manner. System.

The responsibility for activating such a public alert and notification system shall remain with the appropriate Also refer to basis for 10 CFR 50.47(b) and 50.47(b)(10).

44 D.4. If FEMA has approved a nuclear power reactor site's Refer to basis for 10 CFR Part 50, Appendix E, Section IV.D.3 alert and notification design report, including the backup alert regarding the alert and notification system requirements.

and notification capability, as of December 23, 2011, then the backup alert and notification capability requirements in Section IV.D.3 must be implemented by December 24, 2012.

If the alert and notification design report does not include a backup alert and notification capability or needs revision to ensure adequate backup alert and notification capability, then a revision of the alert and notification design report must be

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 23 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption i;:E:MA a1313m .1eEI eaek1::113 aleFt aREI RetifieatieR meaRs m1::1st ee 1

im13lemeRteEI 1Nitl=liR ~e§ Elays afteF FE:MA a1313Feval.

Fle1NeveF, tl=le tetal time 13eFieEI te im13lemeRt a FEMA a1313mveEI eaek1::113 aleFt aREI RetifieatieR meaRs m1::1st Ret e~rneeEI J1::1Re ~~. ~G~ §.

45 E. Emergency Facilities and Equipment No exemption is requested.

Adequate provisions shall be made and described for emergency facilities and equipment, including:

E.1. Equipment at the site for personnel monitoring; 46 E.2. Equipment for determining the magnitude of and for No exemption is requested.

continuously assessing the impact of the release of radioactive materials to the environment; 47 E.3. Facilities and supplies at the site for decontamination of No exemption is requested.

onsite individuals; 48 E.4. Facilities and medical supplies at the site for appropriate No exemption is requested.

emergencv first aid treatment; 49 E.5. Arrangements for medical service providers qualified to No exemption is requested.

handle radioloqical emerqencies onsite; 50 E.6. Arrangements for transportation of contaminated injured No exemption is requested.

individuals from the site to specifically identified treatment facilities outside the site boundary; 51 E. 7. Arrangements for treatment of individuals injured in No exemption is requested.

support of licensed activities on the site at treatment facilities outside the site boundary; 52 E.8.a. (i) A licensee eRsite teel=lRieal s1::11313ert eeRteF aREI aR NEDA analyses demonstrate that 10 months after permanent emeFgeRey e13emtieRs facility from which effective direction cessation of power operations, no remaining postulated accidents at can be given and effective control can be exercised during an DAEC will result in radiological releases requiring offsite protective emergency; actions, or in the event of beyond design basis accidents, 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is available to take mitiQative actions, and if needed, implement offsite

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 24 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption protective actions using a comprehensive emergency management plan. Therefore, there is no need to maintain a TSC or an EOF.

Offsite agency response will not be required at an EOF and onsite actions may be directed from the Control Room or another location, without the requirements imposed on a Technical Support Center (TSC).

An onsite facility will continue to be maintained, from which effective direction can be given and effective control may be exercised during an emergency. The DAEC Emergency Plan will continue to maintain arrangements for requesting assistance and using resources from appropriate offsite support organizations.

53 E.8.a (ii) FSF R1::1eleaF 13e1.veF FeaeteF lieeRsees, a lieeRsee NUREG-0696, "Functional Criteria for Emergency Response Facilities,"

eRsite e13emtieRal s1::11313ert eeRteF; (Reference 21) provides that the operational support center (OSC) is an onsite area separate from the control room and the TSC where licensee operations support personnel will assemble in an emergency.

For a defueled power plant, an OSC is no longer required to meet its original purpose of an assembly area for plant logistical support during an emergency. A single onsite facility will continue to be maintained at DAEC, from which control room support, emergency mitigation, radiation monitoring, and effective control may be exercised during an emen:1encv.

54 E.8.b. F9F a R1::1eleaF 13ev.ieF FeaeteF lieeRsee's eFReF§eRey In accordance with paragraph 8.e. the requirements of paragraph e13eFatieRs faeility FeEJl::liFeel 13y 13aFa§Fa131=! g_a ef tl=!is seetieR, 8.b.(1)- (5) do not apply to the DAEC EOF because it was an eitl=leF a faeility leeateel 13et>.veeR ~ Q FRiles aRel 2§ FRiles ef tl=!e approved facility prior to December 23, 2011. However, the R1::1eleaF 13eweF FeaeteF site~s~, SF a 13FiFRaFy faeility leeateel less exemption is requested to clearly reflect that the requirement no tl=!aR ~ Q FRiles fFeFR tl=!e R1::1eleaF 13eweF FeaeteF site(s~ aRel a longer applies to DAEC in a permanently shut down and defueled l3aek1::113 faeility leeateel 13etweeR ~ Q FRiles aRel 2§ FRiles ef tl=!e condition.

R1::1eleaF 13eweF FeaeteF site(s~. AR eFReF§eRey e13eFatieRs faeility FRay seFve FReFe tl=!aR eRe Rl::leleaF 13eweF FeaeteF site. Also refer to basis for 10 CFR 50.47(b)(3).

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Duane Arnold Energy Center \Attachment 1 to Enclosure of NG-20-0069 I Page 25 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption request prior Commission approval by submitting an application for an amendment to its license. For an emergency operations facility located more than 25 miles from a nuclear power reactor site, provisions must be made for locating NRG and offsite responders closer to the nuclear power reactor site so that NRG and offsite responders can interact face to face with emergency response personnel entering and leaving the nuclear pmver reactor site.

Provisions for locating NRG and offsite responders closer to a nuclear power reactor site that is more than 25 miles from 55 56 57 58 59 60 E.8.c. By June 20, 2012, for a nuclear pmver reactor Refer to basis for 10 CFR Part 50, Appendix E, Section IV.E.8.a(i) and licensee's emergency operations facility required by 10 CFR 50.47(b)(3).

paragraph 8.a of this section, a facility having the following capabilities:

(1) The capability for obtaining and displaying plant data and radiological information for each reactor at a nuclear pmver reactor site and for each nuclear power reactor site that the 61

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 26 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # I 10 CFR Part 50, Appendix E Basis for Exemption 62 I E.8.c (3) The capability to support response to events occurring simultaneously at more than one nuclear power reactor site if the emergency operations facility serves more than one site; and 63 E.8.d. For nuclear povver reactor licensees, an alternative Refer to basis for 10 CFR Part 50, Appendix E, Section IV.1. regarding facility (or facilities) that 'Nould be accessible even if the site hostile action.

is under threat of or experiencing hostile action, to function as a staging area for augmentation of emergency response staff and collectively having the follmNing characteristics: the capability for communication .vith the emergency operations 1

facility, control room, and plant security; the capability to perform offsite notifications; and the capability for engineering assessment activities, including damage control team planning and preparation, for use \Nhen onsite emergency facilities cannot be safely accessed during hostile action. The requirements in this paragraph 8.d must be implemented no later than December 23, 2014, '.vith the exception of the capability for staging emergency response organization personnel at the alternative facility (or facilities) and the capability for communications .vith the emergency 1

operations facility, control room, and plant security, which 64 Refer to basis for 10 CFR 50.47(b)(3) and Appendix E,Section IV.E.8.b.

65 E.9. At least one onsite and one offsite communications Refer to basis for 10 CFR 50.47(b) and 10 CFR 50.47(b)(10).

system; each system shall have a backup power source. All communication plans shall have arrangements for NEDA will maintain communications with the State of Iowa, local emergencies, including titles and alternates for those in government agencies, and the NRC. The onsite response facilities will charae at both ends of the communication links and the be combined into a sinale facilitv as described in IV.E.8.a(ii). A

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 27 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption primary and backup means of communication. Where description of the communications systems and the testing frequency consistent with the function of the governmental agency, will be included in the Permanent Defueled Emergency Plan.

these arrangements will include:

E.9.a. Provision for communications with contiguous State/local governments .vithin the plume exposure pathway 1

~. Such communications shall be tested monthly.

66 E.9.b. Provision for communications with Federal emergency No exemption is requested.

response organizations. Such communications systems shall be tested annuallv.

67 E.9.c. Provision for eommunieations amen§ the nuelear Due to analyses indicting that, within 10 months after shutdown, no power reaetor eontrol room, the onsite teehnieal support credible accident at DAEC will result in radiological releases requiring eenter, anEI the emer§eney operations faeility; anEI amen§ the offsite protective actions, or in the event of beyond design basis nuelear faeility, the prineipal State anEI loeal emer§eney accidents, 1O hours is available to take mitigative actions, and if operations eenters, anEI the fiele assessment teams. Sueh needed, implement offsite protective actions using a comprehensive eommunieations systems shall be testee annually. emergency management plan. Therefore, there is no need for the TSC, EOF, or field assessment teams.

Also refer to basis for 10 CFR 50.47(b)(3).

The provisions remaining in 10 CFR Part 50, Appendix E, Section IV.E.9.a, b, and d include the necessary requirements. Communication with State and local Emergency Operations Centers (EOCs) will be maintained to coordinate assistance on site, if required.

68 E.9.d. Provisions for communications by the licensee with The functions of the control room, EOF, TSC and OSC may be NRC Headquarters and the appropriate NRC Regional Office combined into one or more locations due to the smaller facility staff and Operations Center from the nuelear power reaetor eontrol the greatly reduced required interaction with State and local emergency room, the onsite teehnieal support eenter, ane the emer§eney response facilities. An onsite facility will continue to be maintained, operations facility. Such communications shall be tested from which effective direction can be given and effective control may be monthly. exercised during an emergency. NEDA will maintain communications with the NRC.

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 28 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption Also refer to basis for 10 CFR 50 .4 7 (b).

69 F. Training No exemption is requested.

F.1. The program to provide for: (a) The training of employees and exercising, by periodic drills, of emergency plans to ensure that employees of the licensee are familiar with their specific emergency response duties, and (b) The participation in the training and drills by other persons whose assistance may be needed in the event of a radiological emergency shall be described. This shall include a description of specialized initial training and periodic retraining programs to be provided to each of the following cateqories of emerqencv personnel 70 F.1.i. Directors and/or coordinators of the plant emergency on::ianization; 71 F.1.ii. Personnel responsible for accident assessment, includinq control room shift personnel; 72 F.1.iii Radiological monitoring teams; 73 F.1.iv. Fire control teams (fire briqades);

74 F.1.v. Repair and damaqe control teams; 75 F.1.vi. First aid and rescue teams; 76 F.1.vii. Medical support personnel; 77 F.1.viii. bieeAsee's l=leaElei1::1aFteFs Sl::IFJFJ9Ft FJeFSeAAel; The number of staff at DAEC during the decommissioning process will be small but commensurate with the need to safely store spent fuel at the facility in a manner that is protective of public health and safety.

Excluding public communications, NEDA will maintain a level of emergency response that does not require additional response by headquarters personnel. The on-shift and emergency response positions are defined in the Permanently Defueled Emergency Plan and will be regularly tested through drills and exercises, audited, and inspected by NEDA and the NRC.

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 29 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption Also refer to the basis for 10 CFR 50.47(b). Therefore, exempting licensee's headquarters personnel from training requirements is considered to be reasonable.

78 F.1.ix. Security personnel. No exemption is required.

79 F.1 In addition, a radiological orientation training program Because there will no longer be any expected actions that must be shall be made available to local services personnel; e.g., taken by the public during an emergency, it is no longer necessary to local emergency services/Civil Defense, local law pre-plan the dissemination of this information to the public or to provide enforcement personnel, local news media persons. radiological orientation training to local news media persons.

The phrase "Civil Defense" is no longer a commonly used term and is no longer applicable as an example in the regulation.

80 F.2. The plan shall describe provisions for the conduct of NEDA analyses demonstrate that 10 months after permanent emergency preparedness exercises as follows: cessation of power operations, no remaining postulated accidents at DAEC will result in radiological releases requiring offsite protective Exercises shall test the adequacy of timing and content of actions, or in the event of beyond design basis accidents, 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is implementing procedures and methods, test emergency available to take mitigative actions, and if needed, implement offsite equipment and communications networks, test the public protective actions using a comprehensive emergency management alert and notification system, and ensure that emergency plan. Therefore, the public alert and notification system will not be organization personnel are familiar with their duties. used and no testing would be required.

Also refer to basis for 10 CFR 50.47(b).

81 F.2.a. A full participation e*ercise 1<<1hich tests as Ffluch of the NEDA will continue to invite the State of Iowa and local support licensee, State, and local emer§ency plans as is reasonably organizations to participate in the periodic drills and exercises achievable 1<<1ithout mandatory public participation shall be conducted to assess its ability to perform responsibilities related to an conducted for each site at 1<<1hich a po1<<1er reactor is located. emergency at DAEC to the extent defined by the DAEC Emergency Nuclear po1.ver reactor licensees shall subffiit e*ercise Plan. Because the need for off-site emergency planning is relaxed due scenarios under § 50.4 at least 60 days before use in a full to the low probability of design-basis accidents or other credible events 82

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..... -*-* " - that would be expected to result in an offsite radioactive release that would exceed the limits of EPA PAGs and the available time for event e*ercise must be conducted 1.vithin ti.vo years before the mitigation, no off-site emergency plans will be in place to test.

issuance of the first operatin§ license for full pmver (one

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Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 30 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption first reactor and shall include participation by each State and The intent of submitting exercise scenarios at power reactors is to local government .vithin the plume exposure pathv1ay EPZ 1

check that licensees utilize different scenarios in order to prevent the and each state within the ingestion exposure pathway EPZ. If preconditioning of responders at power reactors. For defueled sites, the full participation exercise is conducted more than 1 year there are limited events that could occur and the previously routine prior to issuance of an operating licensee for full povver, an progression to General Emergency in power reactor site scenarios is exercise which tests the licensee's onsite emergency plans not applicable to a decommissioning site.

must be conducted .vithin one year before issuance of an 1

I I operating license for full power. This exercise need not have I NEDA considers DAEC to be exempt from F.2.a.(i)-(iii) because DAEC State or local government participation. will be exempt from the umbrella provision of Section IV.F.2.a.

83 I F 2.a.(ii) For a combined license issued under part 52 of this chapter, this exercise must be conducted 'Nithin r.vo years of the scheduled date for initial loading of fuel. If the first full participation exercise is conducted more than one year before the scheduled date for initial loading of fuel, an exercise which tests the licensee's onsite emergency plans must be conducted within one year before the scheduled date for initial loading of fuel. This exercise need not have State or local government participation. If FEMA identifies one or more deficiencies in the state of offsite emergency preparedness as the result of the first full participation exercise, or if the Commission finds that the state of emergency preparedness does not provide reasonable assurance that adequate protective measures can and '.Viii be taken in the event of a radiological emergency, the provisions

_r c- c::n C:..A.1--\ ---1, 84 I F 2.a (iii) For a combined license issued under part 52 of this chapter, if the applicant currently has an operating reactor at the site, an exercise, either full or partial participation, shall be conducted for each subsequent reactor constructed on the site. This exercise may be incorporated in the exercise requirements of Sections IV.F.2.b. and c. in this appendix. If i;:i;::~nA irlnntifin!':: nnn nr mnrn rlnfir.innr.in!':: in thn !'::btn nf

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 /Page 31 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption e#site eFReF§eRey 13Fe13aFeEIRess as tt:ie Fes1::1lt ef tt:iis e*eFeise feF tt:ie Rew FeaeteF, eF if tt:ie GeFRFRissieR fiREls tt:iat tt:ie state ef eFReF§eRey 13Fe13aFeEIRess Elees Ret 13FeviEle FeaseRaele ass1::1FaRee tt:iat aEleei1::1ate 13Feteeth1e FReas1::1Fes eaR aREI 11.1ill ee takeR iR tt:ie eveRt ef a mEliele§ieal eFReF§eRey, tt:ie 13rnvisieRs

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85 F 2.b. Each licensee at each site shall conduct a subsequent Refer to basis for 10 CFR Part 50, Appendix E, Section IV.F.2.a.

exercise of its onsite emergency plan every 2 years. ~J1::1eleaF 13e1NeF FeaeteF lieeRsees st:iall s1::1eFRit e*eFeise seeRaFies The low probability of a design-basis accident or other credible events l::IREleF § §Q.4 at least @Q Elays eefeFe 1::1se iR aR e*eFeise that would result in an offsite radioactive release that would exceed the FeEjl::liFeEI ey tl:iis 13aFa§Fa13R 2.13. +Re e*eFeise FRay ee EPA PAGs and the available time for event mitigation at DAEC during iRel1::1EleEI iR tRe f1::1ll 13artiei13atieR eieRRial e*eFeiSe FeEjl::liFeEI ey decommissioning render TSC, OSC and EOF unnecessary. The 13am§FaJ3R 2.e ef tt:iis seetieR. In addition, the licensee shall principal functions required by regulation can be performed at an onsite take actions necessary to ensure that adequate emergency location that does not meet the requirements of the TSC, OSC or EOF.

response capabilities are maintained during the interval The onsite response facilities at DAEC will be combined into a single between biennial exercises by conducting drills, including-at facility.

least one drill involving a combination of some of the principal functional areas of the licensee's onsite emergency response NEDA will continue to conduct biennial exercises and will invite the capabilities. The principal functional areas of emergency State of Iowa and local support organizations (firefighting, law response include activities such as management and enforcement, and ambulance/medical services) to participate in coordination of emergency response, accident assessment, periodic drills and exercises to assess its ability to perform event classification, notification of offsite authorities, responsibilities related to an emergency at DAEC, to the extent defined assessment of the onsite aREI e#site impact of radiological by the DAEC emergency plan.

releases, 13Feteetive aetieR FeeeFRFReRElatieR Elevele13FReRt, 13rnteetive aetieR EleeisieR FRakiR§, 13laRt system repair and mitigative action implementation. During these drills, activation of all of the licensee's emergency response facilities (+eeRRieal S1::11313ert GeRteF (+SG), G13eFatieRs S1::11313ert GeRteF (GSG), aREI tt:ie EFReF§eRey G13emti0Rs Faeility (EGF)) would not be necessary, licensees would have the opportunity to consider accident management strateqies, supervised instruction would be permitted,

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 32 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption operating staff in all participating facilities would have the opportunity to resolve problems (success paths) rather than have controllers intervene, and the drills may focus on the onsite exercise traininq objectives.

86 F 2. c. Offsite plans for each site shall be exercised biennially Refer to basis for 10 CFR Part 50, Appendix E, Section IV.1 and 10 with full participation by each offsite authority having a role CFR Part 50, Appendix E,Section IV.F.2.a.

under the radiological response plan. Where the offsite authority has a role under a radiological response plan for more than one site, it shall fully participate in one exercise every two years and shall, at least, partially participate in other offsite plan exercises in this period. If two different licensees each have licensed facilities located either on the same site or on ~~acent, contiguous sites, and share most of the sRa+J.;. elements defining co located licenses ' then each lice nsee 87 I F 2.c(1) Conduct an exercise bienniallv of its onsite nmnrnnnr.v

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89 I F 2.c.(3) Conduct emergency preparedness activities and interactions in the years ber.veen its participation in the offsite full or partial participation exercise \Nith offsite authorities, to test and maintain interface among the affected State and local authorities and the licensee. Co located licensees shall also participate in emergency preparedness activities and interaction with offsite authorities for the period bet'.veen exercises; 90 I F 2.c.(4) Conduct a hostile action exercise of its onsite emerqencv plan in each exercise cycle; and 91 I F 2.c.(5) Participate in an offsite biennial full or partial participation hostile action exercise in alternating exercise sveles

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 33 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption 92 F 2.d. E:aeR State witR Fes13eAsisility feF A1::1eleaF 13ei11eF Refer to basis for 10 CFR 50.47(b)(10).

FeaeteF emeF§eAey 13Fe13aFeEIAess sRe1::1IEI f1::1lly 13aFtiei13ate iA tRe iA§estieA 13atRv1ay 13eFtieA ef e*eFeises at least eAee evePf e*eFeise eyele. IA States witR meFe tRaA eAe A1::1eleaF 13eweF FeaeteF 13!1::1me e*13es1::1Fe 13atRway E:PZ, tRe State sRe1::1IEI rntate tRis 13aFtiei13atieA fFeFR site te site. E:aeR State witR Fes13eAsisility feF A1::1eleaF 13eweF FeaeteF emeF§eAey 13Fe13aFeEIAess sRe1::1IEI f1::1lly 13aFtiei13ate iA a Restile aetieA e*eFeise at least eAee ei1ePf eyele aAEI sRe1::1IEI f1::1lly 13aFtiei13ate iA eAe Restile aetieA e*eFeise sy QeeemseF dq, ~Wm. States VititR FR9F8 tRaA eAe Al::leleaF 139'.VeF FeaeteF 13l1::1me e*FJ9Sl::lF8

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CD7 ~L..- ,1,..1 --L-L- +h;~ --...L=-=--i=-- "--- ~=i- L- ~=L-93 F 2.e. Licensees shall enable any State or local government Refer to basis for 10 CFR 50.47(b)(10).

leeateEI witRiA tRe 13l1::1FRe e*13es1::1m 13atRway E:PZ to participate in the licensee's drills when requested by such State or local Qovernment.

94 F 2.f. Remedial exercises will be required if the emergency The Federal Emergency Management Agency (FEMA) is responsible plan is not satisfactorily tested during the biennial exercise, for evaluating the adequacy of an offsite response exercise. No action such that NRC, iA eeAs1::1ltatieA witR FE:MA,cannot (1) find is expected from State or local organizations in response to an event at reasonable assurance that adequate protective measures a decommissioning site other than receiving notification of the can and will be taken in the event of a radiological emergency and firefighting, law enforcement and ambulance/medical emergency or (2) determine that the Emergency Response response services. Letters of Agreement will continue to be in place for Organization (ERO) has maintained key skills specific to those services. Offsite response organizations will continue to take emergency response. +Re e*teAt ef State aAEI leeal implement actions to protect the health and safety of the public as they 13aFtiei13atieA iA FeFReElial e*eFeises m1::1st se s1::1#ieieAt te SR9\N would at any other industrial site.

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95 F 2.g. All exercises, drills, and training that provide No exemption is requested.

performance opportunities to develop, maintain, or demonstrate key skills must provide for formal critiques in order to identify weak or deficient areas that need correction.

Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 34 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption Any weaknesses or deficiencies that are identified in a critique of exercises, drills, or training must be corrected.

96 F 2.h. The participation of State and local governments in an No exemption is requested.

emergency exercise is not required to the extent that the applicant has identified those governments as refusing to participate further in emergency planning activities, pursuant to§ 50.47(c)(1 ). In such cases, an exercise shall be held with the applicant or licensee and such governmental entities as elect to participate in the emergency planning process.

97 F 2.i. Licensees shall use drill and exercise scenarios that At DAEC, there will be limited events that could occur that could result provide reasonable assurance that anticipatory responses in radiological releases that exceed the EPA PAGs and the previously will not result from preconditioning of participants. ~ routine progression to General Emergency in power reactor site seeRaFiss feF R1::1eleaF 13s1NeF FeaetsF lieeRsees FR1::1st iRel1::1Ele a scenarios will not be applicable. Therefore, NEDA does not expect to wiEle s13eetF1::1FR sf FaElislsgieal rnleases aREI eveRts, iRel1::1EliRg demonstrate response to a wide spectrum of events.

hsstile aetisR. Exercise and drill scenarios as appropriate must emphasize coordination among onsite and offsite Also refer to the basis for 10 CFR 50.47(b) detailing the low likelihood response organizations. of any credible accident resulting in radiological releases requiring offsite protective measures and refer to the basis for 10 CFR Part 50, Appendix E, Section IV.1 regarding hostile action.

98 F 2.j. +t:ie e*eFeises esREl1::1eteEI 1::1REleF 13aFagFa131=l 2 sf tl=lis Refer to the basis for 10 CFR Part 50, Appendix E, Section IV.F.2.

seetisR 13y Rl::leleaF 13sweF FeaetsF lieeRsees FR1::1st 13FsviEle tl=le s1313sFt1::1Rity feF the E:RG ts EleFRsRstmte 13FsfieieRey iR tl=le l~ey Periodic drills and exercises will be completed to demonstrate ERO skills Reeessai:y ts iFR13leFReRt tl=le 13FiRei13al f1::1RetisRal aFeas sf proficiency in key skills necessary to implement the principal functional eFReFgeRey Fes13sRse iEleRtifieEI iR 13aFagFa13h 2.13 sf this areas of emergency response as applicable for the permanently seetisR. E:aeh e*eFeise FR1::1st 13FsviEle tl=le s1313sFt1::1Rity fsF tl=le defueled plant status. Critiques will follow each drill or exercise activity.

E:RG ts EleFRSRStFate key skills s13eeifie ts eFReFgeRey NEDA will continue to include the State of Iowa and local support Fes13sRse E11::1ties iR the esRtFsl FSSFR, +SG, GSG, E:GF, aREI organizations in the periodic drills and exercises to assess its ability to jsiRt iRfSFFRatisR eeRteF. AEIElitisRally, iR eael=l g ealeRElaF yeaF perform responsibilities related to an emergency at DAEC to the extent e*eFeise eyele, R1::1eleaF 13s11<<eF FeaetsF lieeRsees shall vai:y the defined by the DAEC Emergency Plan.

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Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 35 of 59 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E Basis for Exemption respond to the following scenario elements: hostile action directed at the plant site, no radiological release or an unplanned minimal radiological release that does not require public protective actions, an initial classification of or rapid escalation to a Site Area Emergency or General Emergency, implementation of strategies, procedures, and guidance under§ 50.155(b)(2), and integration of offsite resources with onsite response. The licensee shall maintain a record of exercises conducted during each 8 year exercise cycle that documents the content of scenarios used to comply 1.vith the requirements of this paragraph. Each licensee shall conduct a hostile action exercise for each of its sites no later than December 31, 2015. The first 8 year exercise cycle for a site will begin in the calendar year in which the first hostile action exercise is conducted. For a site licensed under Part 52, the first 8 year exercise cycle begins in the calendar year of the initial exercise required bv Section IV.F.2.a of this appendix.

99 I G. Maintaining Emergency Preparedness I No exemption is requested.

Provisions to be employed to ensure that the emergency plan, its implementing procedures, and emergency equipment and supplies are maintained up to date shall be described.

100 I H. Recovery I No exemption is requested.

Criteria to be used to determine when, following an accident, reentry of the facility would be appropriate or when operation could be resumed shall be described.

101 I I. Onsite Protective Actions During Hostile Action I Refer to the basis for 10 CFR Part 50, Appendix E, Section IV.1.

By June 20, 2012, for nuclear po1.ver reactor licensees, a

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Duane Arnold Energy Center\ Attachment 1 to Enclosure of NG-20-0069 I Page 36 of 59 table 2 Exemptions Requested from 10 CFR 50, Appendix E Item# 10 CFR Part 50, Appendix E I Basis for Exemption hostile action must be developed to ensure the continued ability of the licensee to safely shut dmvn the reactor and

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Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 37 of 59

4. 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 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 7) 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 1) 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 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 1O hours (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.

Table 3 contains a listing of seven analyses that are expected to be evaluated by a decommissioning power reactor licensee requesting exemption of emergency planning requirements. The table also contains a description of how NEDA addresses each of these analyses.

TABLE 3 Interim Staff Guidance-02 Comparison Analysis ISG-02 Description Response 1 Applicable design DBAs (i.e., fuel The postulated accident that will remain handling accident in the spent fuel applicable to DAEC and could contribute to dose storage facility, waste gas system upon implementation of the requested release, and cask handling exemptions is the fuel handling accident (FHA) accident if the cask handling in the Reactor Building, where the SFP is system is not licensed as single- located. The results of the analysis indicate that failure-proof) (Indicates that any the dose at the EAB would not exceed the EPA radiological release would not PAGs within 19 days after permanent cessation exceed the limits of EPA PAGs at of power operations. This analysis is described EAB); in Section 4.A. of this attachment.

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 38 of 59 The Reactor Building Crane is single-failure-proof and therefore a cask handling accident is not credible.

2 Complete loss of SFP water NEDA performed an analysis that conservatively inventory with no heat loss evaluates the length of time (number of hours) it (adiabatic heatup) demonstrating takes for uncovered spent fuel assemblies in the a minimum of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is SFP to reach the temperature at which the available before any fuel cladding zirconium cladding would fail. Based on the temperature reaches 900 degrees limiting fuel assembly for decay heat and Celsius from the time all cooling is adiabatic heat up analysis, at 10 months after lost (Demonstrates sufficient time permanent cessation of power operations, the to mitigate events that could lead time for the hottest fuel assembly to reach to a zirconium cladding fire); 900°C is >1 O hours after spent fuel is uncovered.

This analysis is described in Section 4.B. of this attachment and is included in Attachment 2.

3 Loss of SFP water inventory NEDA performed an analysis to determine the resulting in radiation exposure at offsite radiological impact of a complete loss of the EAB and control room; SFP water. It was determined that 10 months (Indicates that any release is less after shutdown, the gamma radiation dose rate than EPA PAGs at EAB); and at the EAB would be limited to small fractions of the EPA PAG exposure levels.

This analysis is described in Section 4.C. of this attachment and is included in Attachment 3.

4 Considering the site-specific The DAEC SFP is designed as a Seismic seismic hazard, either an Category I structure (UFSAR 9.1.2.3.3.1 ), i.e.,

evaluation demonstrating a high designed to withstand a Safe Shutdown confidence of a low-probability earthquake (SSE) (UFSAR 3.8.2.3.1 ).

(less than 1 x 10-5 per year) of seismic failure of the spent fuel NEDA conducted a seismic evaluation in storage pool structure or an response to Recommendation 2.1 of the Near analysis demonstrating the fuel Term Task Force (NTTF) review of the accident has decayed sufficiently that at the Fukushima Dia-ichi nuclear facility. This natural air flow in a completely evaluation included the spent fuel pool and was drained pool would maintain submitted to the NRC for review (Reference 23).

peak cladding temperature below This evaluation provides a specific assessment 565 degrees Celsius (the point of of earthquake probabilities versus acceleration incipient cladding damage) for the Duane Arnold, and concludes, regardless (Indicates that any release is less of response spectral frequency, the probability is than EPA PAGs at EAB). less than 2 x 1o-6 /year. The NRC review of this evaluation is documented in References 24 and 25.

5 The analyses and conclusions IDCs and SDAs are addressed in Section D and described in NUREG-1738 are Tables 4 and 5 of this attachment.

predicated on the risk reduction measures identified in the study

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 39 of 59 as Industry Decommissioning Commitments (IDC) and Staff Decommissioning Assumptions (SDA), listed in Tables 4.1-1 and 4.1-2 of that document. The staff should ensure that the licensee has addressed these IDCs and SDAs for the decommissioning site if they are storing fuel in an SFP.

6 Verify that the licensee presents The onsite restoration plans for repair of the a determination that there is SFP cooling system and to provide makeup sufficient resources and water to the SFP are incorporated into DAEC adequately trained personnel procedures and utilize adequately trained on-available on-shift to initiate shift resources for implementation.

mitigative actions within the 10-hour minimum time period that There are multiple ways to initiate mitigative will prevent an offsite radiological actions and add makeup water to the SFP within release that exceeds the EPA the 10-hour minimum time period with or without PAGs at the EAB. entry to the SFP floor.

Refer to SDA-2 in Table 5 of this attachment.

7 Verify that mitigation strategies NEDA maintains procedures and strategies for are consistent with that required the movement of any necessary portable by the Permanently Defueled equipment that will be relied upon for mitigating Technical Specifications or by the loss of SFP water. These mitigative retained license conditions. strategies, addressing events involving a loss of SFP cooling and/or water inventory, include implementation of SFP inventory makeup strategies required under 10 CFR 50.155. These diverse strategies provide defense-in-depth and ample time to provide makeup water or spray to the SFP prior to the onset of zirconium cladding ignition when considering very low probability beyond design basis events affecting the SFP.

Refer to SDA-2 in Table 5 of this attachment.

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, concluded that the dose at the EAB 19 days after

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 40 of 59 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 The analysis in Attachment 2 compares the conditions for the hottest fuel assembly stored in the DAEC fuel pool to a criterion proposed in SECY-99-168 (Reference 10) 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. If the heat up time is greater than 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , then offsite emergency planning involving the plant is not necessary.

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 11 ), 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 0.00278 hours 1.653439e-5 weeks 3.805e-6 months 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, and offsite emergency planning involving facilities is not necessary. '

C. Consequences of Other Analyzed Events The analysis in Attachment 3 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 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months 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.

D. Comparison to NUREG-1738 Industry Decommissioning Commitments and Staff Decommissioning Assumptions

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 41 of 59 NEDA also evaluated the industry decommissioning commitments (IDCs) and staff decommissioning assumptions (SDAs) contained in NUREG-1738 (Reference 11 ). NUREG-1738 contains the results of the NRC staffs evaluation of the potential accident risk in spent fuel pools at decommissioning plants in the United States. The study was undertaken to support development of a risk-informed technical basis for reviewing exemption requests and a regulatory framework for integrated rulemaking. The NRC staff performed analyses and sensitivity studies on evacuation timing to assess the risk significance of relaxed offsite emergency preparedness requirements during decommissioning. The staff based its sensitivity assessment on the guidance in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Reference 12). The staff's analyses and conclusions apply to decommissioning facilities with SFPs that meet the design and operational characteristics assumed in the risk analysis.

The study found that the risk at decommissioning plants is low and well within the Commission's Safety Goals. The risk is low because of the very low likelihood of a zirconium fire (resulting from a postulated irrecoverable loss of SFP cooling water inventory).

The study provided the following assessment:

"The staff found that the event sequences important to risk at decommissioning plants are limited to large earthquakes and cask drop events. For emergency planning (EP) assessments, this is an important difference relative to operating plants where typically a large number of different sequences make significant contributions to risk. Relaxation of offsite EP a few months after shutdown resulted in only a "small change" in risk, consistent with the guidance of RG 1.174. Figures ES-1 and ES-2 [in NUREG-1738]

illustrate this finding. The change in risk due to relaxation of offsite EP is small because the overall risk is low, and because even under current EP requirements, EP was judged to have marginal impact on evacuation effectiveness in the severe earthquakes that dominate SFP risk. All other sequences including cask drops (for which emergency planning is expected to be more effective) are too low in likelihood to have a significant impact on risk. For comparison, at operating reactors, additional risk-significant accidents for which EP is expected to provide dose savings are on the order of 1x10-5 per year, while for decommissioning facilities, the largest contributor for which EP would provide dose savings is about two orders of magnitude lower (cask drop sequence at 2x10-7 per year)."

The Executive Summary in NUREG-1738 states, in part, "the staff's analyses and conclusions apply to decommissioning facilities with SFPs that meet the design and operational characteristics assumed in the risk analysis. These characteristics are identified in the study as IDCs and SDAs. Provisions for confirmation of these characteristics would need to be an integral part of rulemaking." The IDCs and SDAs are listed in Tables 4.1-1 and 4.1-2, respectively, of NUREG-1738. The following tables show how the DAEC SFP meets or compares with each of these IDCs (Table 4) and SDAs (Table 5). Attachment 4 includes a new regulatory commitment to update the DAEC UFSAR with this information.

E. Consequences of a Beyond-Design Basis Earthquake

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 42 of 59 NUREG-1738 (Reference 11) 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 7) at the EAB.

The risk associated with zirconium cladding fire events decreases as the spent fuel ages. When 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 22) 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 (i.e., formal offsite radiological emergency preparedness plan or Comprehensive Emergency Management Plan).

The purpose of NUREG-2161 (Reference 22) 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."

NUREG-2161 also states:

"The study shows the likelihood of a radiological release from the spent fuel pool after the analyzed severe earthquake at the reference plant to be about one time in 1O million years or lower. If a leak and radiological release were to occur, this study shows that individuals cancer fatality risk for a member of the public is several orders of magnitude lower than the Commission's Quantitative Health Objective of two in one million (2 x 10- 6/year). For such a radiological release, this study shows public and environmental effects are generally the same or smaller than earlier studies."

The reference plant for the study (a General Electric Type 4 BWR with a Mark I containment) generated approximately 3500 MWt and the SFP contained 2844 fuel assemblies. DAEC is a General Electric Type 4 BWR with a Mark I containment licensed to generate 1912 MWt.

Following permanent cessation of power operations and transfer of all fuel from the reactor vessel to the SFP, the SFP will contain a maximum of 1818 fuel assemblies. Therefore, the risks and consequences of an event involving the SFP at DAEC are bounded by those in the NUREG-2161 study.

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 43 of 59 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 Dia-lchi nuclear facility. This evaluation included the spent fuel pool and was submitted to the NRC for review (Reference 23). This evaluation provides a specific assessment of earthquake probabilities versus acceleration for the Duane Arnold, 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 24 and 25.

Additionally, NEDA has also included the Reactor Building structure (and Spent Fuel pool) into the Maintenance Rule - Structures Monitoring Program. The requires a validation by walkdown and drawing review that there are no changes or degradation of the equipment, structure and components and is completed every 2 years. This will continue until all fuel is removed from the pool.

F. Conclusion Based on the above, NEDA has demonstrated that no credible accident will result in radiological releases requiring offsite protective actions. Additionally, there is sufficient time, resources and personnel available to initiate mitigative actions that will prevent an offsite release that exceeds EPA PAGs.

Duane Arnold Energy Center I Attachment 1 to Enclosure of NG-20-0069 I Page 44 of 59 Table 4 Industry Decommissioning Commitments (IDCs) Comparison IDC Industry Commitments Response 1 Cask drop analyses will be performed or The DAEC crane design is consistent with this commitment. Heavy load lifts in and around single failure-proof cranes will be in use for the area of the spent fuel pool (SFP) are performed by the Reactor Building Crane. The handling of heavy loads (i.e., phase II of design of the crane is single failure-proof. Therefore, the likelihood of dropping the spent fuel NUREG-0612 will be implemented). casks in and around the SFP is extremely low. The design meets the requirements of NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, and Appendix C of NUREG-0612, Control of Heavy Loads at Nuclear Power Plants. DAEC procedures provide instructions for lifting activities to meet the guidance provided in NUREG-0612.

Because the Reactor Building Crane is single failure-proof, an accidental load drop is considered not to be a credible event such that condition 5.1.2(1) of NUREG-0612 is satisfied and analysis of cask drop accidents in accordance with condition 5.1.2(4) of NUREG-0612 is not required.

2 Procedures and training of personnel will DAEC procedures are in place to ensure onsite and offsite resources can be brought to bear be in place to ensure that onsite and offsite during an event, including the following:

resources can be brought to bear during an

EPIP 1.2, Notifications event.

DAEC Emergency Plan, Section 'O', Radiological Emergency Response Training

DAEC Emergency Plan, Appendix 2, Letters of Agreement

SY-AA-102-1019, Suspension I Relaxation of Safeguards

DAEC Catastrophic Event Plan

SAMP 716, Initial Response Extensive Damage Mitigation Guidelines (EDMG)

OP-AA-107, Extensive Damage Management Program

EN-AA-110, Diverse and Flexible Coping Strategies (FLEX) Program

EMG, Emergency Management Guideline

FLEX-AB-100, DAEC Flex Program The procedures listed above (or equivalent) and associated training will be updated as necessary to reflect the permanently shut down and defueled condition. Once DAEC is permanently shut down and defueled, the on-shift plant operators, including Certified Fuel Handlers (CFHs) and Non-Certified Operators, will be appropriately trained on the relevant procedures and on the various actions needed to provide makeup to the SFP based on a systematic approach to training to ensure appropriate personnel receive initial and continuing training on beyond design basis event strategies required by 10 CFR 50.155. The DAEC CFH training program was submitted for NRC review on January 29, 2019, and approved by the NRC per letter dated August 29, 2019 (References 9 and 13).

Duane Arnold Energy Center I Attachment 1 to Enclosure of NG-20-0069 I Page 45 of 59 Table 4 Industry Decommissioning Commitments (IDCs) Comparison IDC I Industry Commitments Response Following permanent cessation of power operations, maintaining SFP cooling and inventory would be the highest priority activity. Therefore, the personnel needed to perform these actions will be available at all times.

Finally, periodic Emergency Plan drills are conducted with opportunities for participation of the Offsite Response Organizations to maintain proficiency in response to a plant event.

3 Procedures will be in place to establish The following procedures provide guidance for initiating and maintaining communications communication between onsite and offsite between offsite agencies and the onsite Emergency Response Organization during severe organizations during severe weather and weather and seismic events:

seismic events.

EPIP 1.1, Determination of Emergency Action Levels

EPIP 1.2, Notifications

DAEC Plan, Section F, Emergency Communications

AOP 901, Earthquake

AOP 903, Severe Weather 4 An offsite resource plan will be developed FLEX-AB-100-1003, SAFER Response Plan for Duane Arnold Energy Center, contains an which will include access to portable offsite resource list which shows providers, their capabilities, and a contact telephone pumps and emergency power to number.

supplement onsite resources. The plan would principally identify organizations or suppliers where offsite resources could be obtained in a timely manner.

5 SFP instrumentation will include readouts SFP Level instrumentation provides indication and alarm to Control Room. This consists of and alarms in the control room (or where two pool level instruments installed in accordance with NRC Order EA-12-051 and one pool personnel are stationed) for SFP level instrument on the main control board. The SFP is also equipped with a local level temperature, water level, and area indicator (ruler) for alternate means of determining spent fuel pool level.

radiation levels.

SFP System temperature is continuously monitored in the control room.

There are four area radiation monitors on the Refuel Floor that provide remote indication and annunciation in the Control Room. A local alarm to notify personnel of high area radiation levels is also in place. In addition, each radiation monitor provides input to the Plant Process computer.

Duane Arnold Energy Center I Attachment 1 to Enclosure of NG-20-0069 I Page 46 of 59 Table 4 Industry Decommissioning Commitments {IDCs) Comparison IDC Industry Commitments Response 6 SFP seals that could cause leakage The DAEC SFP gate are static seals, and there is no credible catastrophic failure leading to fuel uncovery in the event of mechanism for these seals. If SFP inventory were to leak due to seal rupture or degradation, seal failure shall be self limiting to leakage level would not go below the top of the spent fuel racks. The fixed elevation of the bottom of or otherwise engineered so that drainage the refueling slot between the SFP and reactor vessel where the gates are located is above cannot occur. the top of spent fuel. Therefore, leakage by the gates could not lead to fuel uncovery.

Additionally, DAEC has a flow indicating switch installed to monitor for any leakage past the SFP qates.

7 Procedures or administrative controls to Administrative controls are in place the drive procedure use and adherence, and risk reduce the likelihood of rapid draindown management. Specifically, AD-AA-100-1006, Procedure and Work Instruction Use and events will include (1) prohibitions on the Adherence, establishes the expectations and requirements for procedure adherence and use of pumps that lack adequate siphon usage for all plant personnel performing activities. Additionally, all work activities are subject protection or (2) controls for pump suction to the work process controls per WM-AA-100-1000, Work Activity Risk Management.

and discharge points. The functionality of anti-siphon devices will be periodically DFS-201, Dry Shielded Canister/Transfer Cask Preparation for Fuel Loading Operations, verified. requires the Cask Pool Gate be installed and a specified volume (5800 gallons) be drained from the Cask Pit prior to placing the cask in the pit. The Cask Pool Gate prevents cask evolutions from affecting SFP level and DFS-201 meets the requirements of the IDC by controlling the draining methods to prevent affecting SFP level. There are no connections to the fuel storage pool that could allow the fuel pool to be drained below the pool gate between the reactor well and fuel pool. The return cooling water supply piping terminates just below the surface of the spent fuel pool. That piping contains passive vacuum breaking vent pipes that prevent any siphoning from occurring on the return lines. These vent pipes are easily observable to verify the absence of any obstructions, and they require no testing since there are no moving parts. The suction piping is routed from the Skimmer Surge Tanks that are connected to the SFP via SFP overflow weirs that maintain SFP level at the required level. High- and low-level switches indicate pool water level changes in the control room and pump room.

8 An onsite restoration plan will be in place AOP 435, Loss of Spent Fuel Pool Cooling (All Modes) I Inventory (Mode 4 and 5), SEP to provide repair of the SFP cooling 312, Loss of Spent Fuel Pool Inventory, SEP 314, Loss of Spent Fuel Pool Cooling and systems or to provide access for makeup SAMP 712, Spent Fuel Pool Makeup and Spray, all provide guidance for SFP makeup water to the SFP. The plan will provide for utilizing various water sources, with or without access to the Refuel Floor.

remote alignment of the makeup source to the SFP without requiring entry to the refuel floor.

Duane Arnold Energy Center I Attachment 1 to Enclosure of NG-20-0069 /Page 47 of 59 Table 4 Industry Decommissioning Commitments (IDCs) Comparison IDC Industry Commitments Response 9 Procedures will be in place to control SFP Administrative controls are in place the drive procedure use and adherence, and risk operations that have the potential to rapidly management. Specifically, AD-AA-100-1006, Procedure and Work Instruction Use and decrease SFP inventory. These Adherence, establishes the expectations and requirements for procedure adherence and administrative controls may require usage for all plant personnel performing activities. Additionally, all work activities are subject additional operations or management to the work process controls per WM-AA-100-1000, Work Activity Risk Management.

review, management physical presence for designated operations or administrative DFS-201 and 01 435, Fuel Pool Cooling System, both require the cask pool gate be limitations such as restrictions on heavy installed prior to draining inventory from the cask pool. The cask pool gate prevents cask load movements. evolutions from affecting SFP level, and DFS-201 and 01 435 meet the requirements of the IDC by controlling the draining methods to prevent affecting SFP level.

Movement of the Dry Storage Cask and other heavy loads in the vicinity of the SFP is performed in accordance with PDA procedure ACP 1408.19, Control of Generic Heavy Loads, and DFS-201 which ensure the requirements of NUREG-0612 are met for heavy loads.

10 Routine testing of the alternative fuel pool AOP 435 lists the makeup sources to use in the event of a loss of SFP inventory, with' or makeup system components will be without access to the Refueling Floor (51h Floor of the Reactor Building). Various systems performed and administrative controls for used in that procedure are either routinely used or tested in accordance with Technical equipment out of service will be Specification or other administrative requirements. For instance, makeup to the Skimmer implemented to provide added assurance Surge tanks via Condensate Service Water is utilized on a daily basis to make up for that the components would be available, if evaporative loses from the SFP. In addition, SFP Cooling risk and equipment out of service needed. times are managed in accordance with OP-AA-104-1010, Spent Fuel pool Risk Manaaement.

Duane Arnold Energy Center I Attachment 1 to Enclosure of NG-20-0069 I Page 48 of 59 Table 5 Staff Decommissioning Assumptions (SDAs) Comparison SDA Staff Assumptions Response 1 Licensee's SFP cooling design will be at The DAEC design aligns with the intent of this description. The DAEC SFP is designed as a least as capable as that assumed in the Seismic Category I structure (UFSAR 9.1.2.3.3.1), i.e., designed to withstand a Safe risk assessment, including instrumentation. Shutdown earthquake (SSE) (UFSAR 3.8.2.3.1). The SFP cooling system is as originally Licensees will have at least one motor- designed and does not include temporary configurations which would result in loss of driven and one diesel-driven fire pump margin or unanalyzed drain paths.

capable of delivering inventory to the SFP.

The instrumentation includes the Fukushima Lessons Learned dual, independent level monitors with indicators and alarms in the Control Room. Temperature indication and alarms are available.

The SFP cooling system has redundant pumps, redundant heat exchangers and multiple makeup sources, including the Fire Protection System. DAEC's Fire Protection System includes an electric-driven fire pump and a diesel-driven fire pump, both of which will be maintained until all fuel is removed from the SFP. Each fire pump has the capability to deliver 500 gallons per minute (gpm) of makeup water to the SFP. All sources discussed above take suction from the Cedar River.

2 Walk-downs of SFP systems will be DAEC performs in-plant walk-downs of the Refuel Floor and SFP demineralizer instruments performed at least once per shift by the for system pressure and flow (which will indicate a system problem) shiftly, as directed by operators. Procedures will be developed Operator rounds. Additionally, in-plant skimmer surge tank level and SFP pump checks are for and employed by the operators to performed once per day as directed by Operator rounds. The daily monitoring of skimmer provide guidance on the capability and surge tank level and SFP pumps are adequate since the continuous indication of SFP level availability of onsite and offsite inventory in the Main Control Room and shiftly monitoring of Refuel Floor and demineralizers will makeup sources and time available to indicate any problems with SFP system operation.

initiate these sources for various loss of cooling or inventory events. DAEC procedures meet the requirements of this SDA by providing the guidance on the capability and availability of permanent and portable makeup sources. AOP 901 directs the inspection of the SFP and cooling systems following a seismic event. AOP 435 includes methods to diagnose the loss of cooling and/or inventory and direction to establish makeup.

SAMP 712, Spent Fuel Pool Makeup and Spray, provides direction in a Beyond Design-Basis External Event (BDBEE).

NEDA has determined that for a loss of SFP cooling with no makeup capabilities, the total time to boil the SFP and to reduce SFP water inventory to a point 10 feet above the top of the highest point of the fuel assembly is calculated to be 2.8 days total. The 2.8-day period is based on the expected decay heat load following a 90-day period following reactor

Duane Arnold Energy Center I Attachment 1 to Enclosure of NG-20-0069 I Page 49 of 59 Table 5 Staff Decommissioning Assumptions (SDAs) Comparison SDA Staff Assumptions Response shutdown. NEDA has performed an evaluation demonstrating that 90 days after permanent shutdown, adequate time and water resources will be available to restore spent fuel pool cooling to maintain SFP water level 10 feet about the top of the spent fuel and the low decay heat and long time to boil off inventory provides sufficient time for DAEC to sustain the SFP cooling function indefinitely. Specifically, DAEC's standard portable fire pumps deliver adequate head and flow to provide the minimum require makeup to the SFP. The equipment can be installed by the Technical Specifications' minimum required staffing within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to deliver SFP makeup. The required equipment and installation procedures are required to be maintained per DAEC Operating License section 2.C.(9), "Mitigation Strategy License Condition," item (b)(7).

3 Control room instrumentation that monitors Indication for SFP level is provided in the Control Room as well as locally in the plant. A SFP temperature and water level will Control Room annunciator will actuate when level is low or level is high. Additionally, if a directly measure the parameters involved. high or low level condition exists, an alarm light (red) will illuminate on local panels in the Level instrumentation will provide alarms at plant. Two independent indicators for SFP level are also provided on the Control Room back levels associated with calling in offsite panels for use during a Beyond Design Basis External Event (BDBEE).

resources and with declaring a general emergency. A temperature element on the common suction to the fuel pool cooling pumps provides temperature indication to recorder TRS 1945, which is located in the Control Room.

NEDA has procedures in place to respond to an abnormally low level in the SFP to direct the plant staff to take appropriate actions to provide the necessary SFP makeup; first through normal means, then by utilizing all available onsite resources, including both design basis and defense-in-depth capabilities. Refer to the NEDA responses for IDC 2 and IDC 4 for details associated with calling in offsite resources.

Regarding the declaration of a general emergency, NEDA will be employing Shutdown EALs using an approved NRC EAL Scheme. Based on Appendix C of NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," Revision 6, it is expected that station conditions will not have the capacity to reach any threshold requiring the declaration of a qeneral emerqencv.

4 Licensee determines that there are no The DAEC SFP Cooling System has not been modified from the original design in order to drain paths in the SFP that could lower the enter into the decommissioning process.

pool level (by draining, suction, or pumping) more than 15 feet below the The normal pool operating level is elevation (EL) 853'-8". Top of active fuel installed in the normal pool operatinq level and that fuel storaqe rack is at EL 830'-3". The water in the SFP returns to the SPF coolinq system

Duane Arnold Energy Center I Attachment 1 to Enclosure of NG-20-0069 I Page 50 of 59 Table 5 Staff Decommissioning Assumptions (SDAs) Comparison SDA Staff Assumptions Response licensee must initiate recovery using offsite via a skimmer weir that can be set to maintain SFP level as low as EL 853'-6.5". There are sources. no lower elevation piping penetrations in the SFP. The bottom of the fuel transfer gate connecting the SFP to the reactor cavity is at EL 831'-2.75". The bottom of the cask pool gate opening is EL 832'-3". The SFP cooling inventory is normally supplied through two 6" pipes that discharge into the SFP at EL 850'. These discharge lines each include a%" high-point vent (welded at EL 852'-6" and open to atmosphere at EL 853'-3"), which act as vacuum breakers to prevent siphoning of the pool through the primary make-up piping.

Therefore, although draining of more than 15 feet below normal pool operating level could occur, there is no drain path that would drain water below the top of the fuel.

NEDA maintains procedures and guidelines in place to obtain offsite assistance if necessary for mitigation of events that result in significant loss of SFP inventory. These mitigating strategies are implemented as part of AOP-435 and are also included in DAEC's Mitigation Strateav License Condition requirements.

5 Load Drop consequence analyses will be Heavy load lifts in and around the area of the SFP are performed by the Reactor Building performed for facilities with nonsingle Crane. The design of the Reactor Building Crane is single failure-proof as noted in response failure-proof syste_ms. The analyses and to IDC-1. Therefore, performance of load drop consequence analyses is not required.

any mitigative actions necessary to preclude catastrophic damage to the SFP that would lead to a rapid pool draining would be sufficient to demonstrate that there is high confidence in the facilities ability to withstand a heavy load drop.

6 Each decommissioning plant will NEDA conducted a seismic evaluation in response to Recommendation 2.1 of the Near successfully complete the seismic checklist Term Task Force (NTTF) review of the accident at the Fukushima Dia-ichi nuclear facility.

provided in Appendix 2B to this study This evaluation included the spent fuel pool and was submitted to the NRC for review

[NUREG-1738]. If the checklist cannot be (Reference 23). This evaluation provides a specific assessment of earthquake probabilities successfully completed, the versus acceleration for the Duane Arnold, and concludes, regardless of response spectral decommissioning plant will perform a plant frequency, the probability is less than 2 x 1o-6/year. The NRC review of this evaluation is specific seismic risk assessment of the documented in References 24 and 25.

SFP and demonstrate that SFP seismically induced structural failure and rapid loss of Additionally, NEDA has also included the Reactor Building structure (and Spent Fuel pool) inventory is less than the generic bounding into the Maintenance Rule - Structures Monitoring Program. The requires a validation by estimates provided in this study (<1 x10- 5 walkdown and drawing review that there are no changes or degradation of the equipment, per year includina non-seismic events).

Duane Arnold Energy Center I Attachment 1 to Enclosure of NG-20-0069 I Page 51 of 59 Table 5 Staff Decommissioning Assumptions (SDAs) Comparison SDA Staff Assumptions Response structure and components and is completed every 2 years. This will continue until all fuel is removed from the pool.

In addition, as documented in the Enhanced Seismic Checklist (NUREG 1738 Appendix 28, Attachment 1), risks associated with a seismic event are mitigated by delaying any EP rule changes until after the zirc fire time period. As this amendment will not be implemented until after the zirc fire period (10 months), overall risk is reduced even further.

7 Licensees will maintain a program to The DAEC spent fuel racks utilize Boral, rather than Boraflex, as the neutron absorbing provide surveillance and monitoring of material. As described in Section 18.1.41 of the DAEC UFSAR, an aging management Boraflex in high-density spent fuel racks program is in place to manage loss of material and reduction of neutron absorption capacity until such time as spent fuel is no longer of Boral neutron absorption panels in the spent fuel racks. The loss of material and the stored in these high-density racks. reduction of the neutron-absorbing capacity will be determined through coupon testing for Holtec spent fuel racks and in situ testing for PaR spent fuel racks. Such testing will include periodic verification of boron loss through areal density measurement of coupons or through direct in situ techniques, such as measurement of boron areal density and measurement of geometric changes in the material, and detection of gaps through blackness testing.

Duane Arnold Energy Center Attachment 1 to Enclosure of NG-20-0069, Page 52 of 59 V. 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. Therefore, the exemption is authorized by law.

B. The exemptions will not present an undue risk to public health and safety The underlying purpose of 10 CFR 50.47(b), 10 CFR 50.47(c)(2), 10 CFR Part 50, Appendix E, Section IV is to ensure that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, to establish plume exposure and ingestion pathway emergency planning zones for nuclear power plants, and to ensure that licensees maintain effective offsite and onsite emergency plans.

As discussed in this request, 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 10 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.

For these reasons, offsite emergency response plans will no longer be needed for protection of the public beyond the EAB 10 months after permanent cessation of power operations. Based on the reduced consequences of radiological events possible at DAEC in the permanently defueled condition, the scope of the onsite emergency preparedness organization and corresponding offsite requirements in the emergency plan may be accordingly reduced without an undue risk to the public health and safety.

Therefore, the underlying purpose of the regulations will continue to be met. Because the underlying purpose of the rules will continue to be met, the exemptions will not present an undue risk to the public health and safety.

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 53 of 59 C. The exemptions are consistent with the common defense and security The reduced consequences of radiological events that will remain possible at the site once it is in the permanently defueled condition allows for a corresponding reduction in the scope of the onsite emergency preparedness organization and associated reduction of requirements in the emergency plan. These reductions will not adversely affect DAEC'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 shutdown 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 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR Part 50, Appendix E is to ensure that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, to establish plume exposure and ingestion pathway emergency planning zones for nuclear power plants, and to ensure that licensees maintain effective offsite and onsite emergency plans.

The standards and requirements in 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR Part 50, Appendix E were developed taking into consideration the risks associated with operation of a nuclear power reactor at its licensed full power level.

These risks include the potential for a reactor accident with offsite radiological dose consequences.

The radiological consequences of accidents that will remain possible at DAEC upon permanent shut down of power operations are substantially lower than those at an operating plant. The upper bounds of the analyzed dose consequences limits the highest attainable emergency class to the Alert level. In addition, because of the reduced consequences of radiological events that will still be possible at the site, the scope of the onsite emergency preparedness organization may be reduced accordingly. Thus, the underlying purpose of the regulations will not be adversely affected by eliminating offsite emergency planning activities or reducing the scope of onsite emergency planning as described in this request.

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 54 of 59 Radiological analysis indicates that within 19 days after shutdown, the radiological consequences of the postulated accident that will remain possible at DAEC upon permanent removal of fuel from the reactor will not exceed the limits of the EPA PAGs at the EAB. In addition, an analysis has been developed for beyond design basis events related to the SFP which show that within 10 months after permanent cessation of power operations, 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 PAGs at the EAB. Therefore, application of all of the standards and requirements in 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR Part 50, Appendix E are not necessary to achieve the underlying purpose of those rules.

Since the underlying purposes of the rules would continue to be achieved even with NEDA being permitted to reduce the scope of emergency preparedness requirements consistent with placing the facility in the permanently defueled condition, application of the rules is not necessary to achieve the underlying purpose, and the 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, or that are significantly in excess of those incurred by others similarly situated. (10 CFR 50.12(a)(2)(iii))

Application of all of the standards and requirements in 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR Part 50, Appendix Eis not needed for adequate emergency response capability and is excessive for a permanently shut down and defueled facility. Application of all of these standards and requirements would result in undue costs being incurred for the maintenance of an emergency response organization in excess of that actually needed to respond to the diminished scope of credible events.

Other licensees, similarly situated, have been granted similar exemptions as discussed in Section E.

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.

3. The exemptions would result in benefit to the public health and safety that compensates for any decrease in safety that may result from the grant of the exemptions. (10 CFR 50.12(a)(2)(iv))

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 permanently shut down and defueled, the postulated accidents that could occur during reactor operation will no longer be possible. Specifically, 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.

The proposed exemptions would allow NEDA to revise the onsite emergency plan to correspond to the reduced scope and consequences of remaining accidents and

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 55 of 59 events. As such, the emergency plan would no longer need to address response actions for events that would no longer be possible. The revised emergency plan would thereby enhance the ability of the emergency response organization to respond to those scenarios that remain credible because emergency preparedness training and drills would focus only on applicable activities. Elimination of requirements for classification of emergency action levels for events that were no longer possible would enhance the ability of the ERO to correctly classify those events that remain credible. As the proposed exemptions will enhance the ability of the organization to respond to credible events, a resultant benefit to the public health and safety is realized.

Therefore, since granting the exemptions would result in benefit to the public health and safety and would not result in a decrease in safety, the special circumstances required by 10 CFR 50.12(a)(2)(iv) exist.

E. Precedent The exemption requests for 10 CFR 50.47(b); 10 CFR 50.47(c)(2); and 10 CFR Part 50, Appendix E requirements are consistent with exemptions on the same emergency planning requirements that recently have been issued by the NRC for other nuclear power reactor facilities beginning decommissioning. Specifically, the NRC granted similar exemptions to OPPD for FCS (Reference 14) to ENO for VY (Reference 15); to Southern California Edison Company for SONGS, Units 1, 2, and 3 (Reference 16); to Duke Energy Florida, Inc. for CR3 (Reference 17); to Dominion Energy Kewaunee, Inc. for KPS (Reference 18); and to ENO for PNSP (Reference 19). Similar to the current request, these precedents each resulted in exemptions from certain emergency planning requirements in 10 CFR 50.47(b); 10 CFR 50.47(c)(2); and 10 CFR Part 50, Appendix E, related to the elimination of offsite radiological emergency plans and reduction in the scope of the onsite emergency planning activities. For the same reasons that the NRC recently issued these exemptions, NEDA seeks approval of the enclosed proposed exemption requests.

VI. 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 requirements of an administrative, managerial, or organizational nature. 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 portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR Part 50, Appendix E would allow NEDA to revise the scope of the DAEC

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 56 of 59 Emergency Plan to reflect the permanently shut down and defueled condition of the station. NEDA has evaluated the proposed exemption 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 exemptions have 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 exemptions 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 accident.

When the exemptions become 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 exemptions.

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

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

The proposed exemptions do 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 exemptions 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. Do the proposed exemptions involve a significant reduction in a margin of safety?

The proposed exemptions do 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.

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 57 of 59 Therefore, the proposed exemptions do not involve a significant reduction in a margin of safety.

Based on the above, NEDA concludes that the proposed exemptions 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 exemptions. 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 exemptions. The proposed exemptions 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 exemptions 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) Requirements of an administrative, managerial, or organizational nature.

The proposed exemptions will form the basis for a reduction in size of the NEDA emergency response organization commensurate with the reduction in consequences of radiological events that will be possible at DAEC once the facility is in the permanently defueled condition. They will also modify the requirements for emergency planning.

Therefore, the exemptions address requirements of an administrative, managerial, or organizational nature.

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 58 of 59 VII. REFERENCES

1. NSIR/DPR-ISG-02, Interim Staff Guidance, Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants

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

3. Federal Register Notice (60 FR 32430) Final rule - 10 CFR Part 72, "Emergency Planning Licensing Requirements for Independent Spent Fuel Storage Facilities (ISFSI) and Monitored Retrievable Storage Facilities (MRS)," dated June 22, 1995 (ML072910459)

4. NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," Revision 6

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

Commission Paper SECY-00-0145, June 28, 2000 (ML003721626)

6. Letter from USN RC to NEI, "U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, Dated November 2012 (TAC NO. D92368)," dated March 28, 2013 (ML12346A463)

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

8. Federal Register Notice, Vol. 76, No. 226 (76 FR 72596), Enhancements to Emergency Preparedness Regulations, dated November 23, 2011

9. Letter from NEDA (D. Curtland) to USNRC, "Request for Approval of Certified Fuel Handler Training Program," NG-19-0003, dated January 29, 2019(ML19037A016)

10. Commissioning Paper SECY-99-168, Improving Decommissioning Regulations for Nuclear Power Plants, dated June 30, 1999

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

12. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated July 1998 (ML003740133)

13. Letter from USNRC to NEDA (D. Moul), "Duane Arnold Energy Center - Approval of a Certified Fuel Handler Training and Continuing Training Program (EPID L-20190LLL-0003),"

dated August 28, 2019 (ML19204A287)

14. Letter, USNRC to OPPD (M. Fisher), "Fort Calhoun Station, Unit No. 1 - Exemptions from Certain Emergency Planning Requirements and Related Safety Evaluation (CAC No.

MF9067; EPID L2016-LLE-0003)," dated December 11, 2017(ML172638198)

15. Letter, USNRC to ENO, "Vermont Yankee Nuclear Power Station - Exemptions from Certain Emergency Planning Requirements and Related Safety Evaluation (CAC No. MF3614),"

dated December 10, 2015 (ML15180A054)

16. Letter from USNRC to Southern California Edison Company (T. Palmisano), "San Onofre Nuclear Generating Station, Units 1, 2, and 3 and Independent Spent Fuel Storage Installation - Exemptions from Certain Emergency Planning Requirements and Related Safety Evaluation (TAC Nos. MF3835, MF3836, and MF3837)," dated June 4, 2015.

(ML15082A204)

17. Letter from USN RC to Crystal River Nuclear Plant (T. Hobbs), "Crystal River Unit 3 -

Exemptions from Certain Emergency Planning Requirements and Related Safety Evaluation (TAC No. MF2981)," dated March 30, 2015. (ML15058A906)

18. Letter from USNRC to Dominion Energy Kewaunee, Inc. (D. Heacock), "Kewaunee Power Station - Exemptions from Certain Emergency Planning Requirements and Related Safety Evaluation (TAC No. MF2567)," dated December 27, 2014. (ML14261A223)

19. Letter from USNRC to Holtec Decommissioning International, LLC (P. Cowan), "Pilgrim Nuclear Power Station - Exemptions from Certain Emergency Planning Requirements and

Duane Arnold Energy Center to Enclosure of NG-20-0069, Page 59 of 59 Related Safety Evaluation (EPID L-2018-LLE-0011 ), " dated December 18, 2019.

(ML19142A043)

20. Federal Register Notice, Vol. 74, No. 94 (74 FR 23254), Enhancements to Emergency Preparedness Regulations, dated May 18, 2009

21. NUREG-0696, "Functional Criteria for Emergency Response Facilities," February 1981

22. 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)

23. 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)

24. 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)

25. Letter from USN RC to Listed Power Reactor Licensees, "Staff Review of High Frequency Confirmation Associated with Reevaluated Seismic Hazard lmplementating Near-Term Task Force Recommendation 2.1 "Seismic" for Specific Licensees," dated February 18, 2016, (ML15364A544)

v t e Letter Sequence Response to RAI
EPID:L-2020-LLA-0112, NextEra Energy Duane Arnold, LLC - Request for Rescission of Security Orders (Open)
Initiation Request, Request, Request Acceptance... Supplement Administration Questions Answers 7 October 2020 Results Approval... Other: ML21067A642
MONTHYEAR2020-08-27 [Table View]

NG-20-0069, Response to Request for Additional Information Relating to Request for Exemption from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E

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NG-20-0069, Response to Request for Additional Information Relating to Request for Exemption from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E

Text

Subject:

References:

2. Background

Navigation menu

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