Text

Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E
	ML19182A104
Person / Time
Site:	Three Mile Island
Issue date:	07/01/2019
From:	Gallagher M; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TMl-19-040
	Download: ML19182A104 (132)
	v • d • e

Michael P. Gallagher Exelon Nucl ear Exelon Generation . Vice Presi dent License Renewal and Oecomm1ss1on1ng 200 Exelon Way Kennett Square, PA 19348 610 765 5958 Office 610 765 5658 Fax www.exeloncorp.com michaelp.gallagher@exeloncorp.com 10 CFR 50. 12 10 CFR 50.47 10 CFR 50, Appendix E TMl-19-040 July 1, 2019 U.S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289 Three Mile Island Nuclear Station, Unit 2 Possession Only License No. DPR-73 NRC Docket No. 50-320

Subject:

Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E

Reference:

1. Letter from J. Bradley Fewell (Exelon Generation Company, LLC) to U.S.

Nuclear Regulatory Commission, "Certification of Permanent Cessation of Power Operations for Three Mile Island Nuclear Station, Unit 1," dated June 20, 2017(ML17171A151)

2. Letter from U.S. Nuclear Regulatory Commission to Bryan C. Hanson, (Exelon Generation Company, LLC}, "Three Mile Island Nuclear Station, Units 1 and 2 - Issuance of Amendment No. 296 for Unit 1 RE: Changes to Emergency Plan for Post-Shutdown and Permanently Defueled Condition (EPID L-2018-LLA-0073}, dated April 18, 2019 (ML19065A114)

Pursuant to 10 CFR 50.12, "Specific exemptions," Exelon Generation Company, LLC (Exelon) requests exemptions from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR Part 50, Appendix E for Three Mile Island Nuclear Station (TMI). The requested exemptions would allow TMI to reduce emergency planning requirements consistent with the permanently defueled condition of the station.

By letter dated June 20, 2017 (Reference 1), Exelon provided formal notification to the U.S.

Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.4(b)(8) and 10 CFR 50.82(a)(1 )(i) of Exelon's determination to permanently cease operations at TMI, Unit 1 (TMl-1) on or about September 30, 2019.

U.S. Nuclear Regulatory Commission TMI Request for Exemption Docket Nos. 50-289 and 50-320 July 1, 2019 Page 2 Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel are submitted to the NRG pursuant to 10 CFR 50.82(a)(1)(i) and (ii),

and pursuant to 10 CFR 50.82(a)(2), the 10 CFR 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel.

By letter April 18, 2019 (Reference 2), the NRG issued the approved changes to the TMI site emergency plan (SEP) to support the planned permanent cessation of operation and permanent defueling at the TMl-1 reactor. The approved changes revise the TMI SEP emergency response organization (ERO) on-shift and augmented staffing, commensurate with the reduced spectrum of credible accidents for a permanently shutdown and defueled nuclear power reactor facility.

Three Mile Island, Unit 2 (TMl-2), has a possession only license and is currently maintained in accordance with the NRG approved SAFSTOR condition (method in which a nuclear facility is placed and maintained in a condition that allows it to be safely stored and subsequently de-contaminated) known as Post-Defueling Monitored Storage (PDMS). Exelon maintains the emergency planning responsibilities for TMl-2, which is owned by First Energy Corporation, through a service agreement. This request for exemptions does not impact Exelon's ability to maintain the service agreement.

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

More specifically, application of the portions of the regulations from which exemptions are sought is not necessary to ensure adequate emergency response capability for TMI and to achieve the underlying purpose of the rules. Furthermore, continued application of these portions of the regulations from which exemptions are sought would result in an undue hardship or other costs to the TMl-1 Decommissioning Trust Fund by requiring continued implementation of unnecessary emergency response capabilities. Finally, granting the requested exemptions would result in benefit to the public health and safety and would not result in a decrease in safety, because they would enhance the ability of the emergency response organization to respond to credible scenarios.

The exemption requests are contained in Attachment 1 to this letter. Exelon has performed analyses which show that 488 days after permanent cessation of power operations, the spent fuel stored in the spent fuel pool will have decayed to the extent that the requested exemptions may be implemented at TMl-1. Following the TMl-1 shutdown, which is expected by the end of September 2019 (Reference 1), 488 days after shutdown is expected to be about January 30, 2021. The bounding analysis is contained in Attachment 2.

TMl-1 plans to submit a Permanently Defueled Emergency Plan (PDEP), containing a Permanently Defueled Emergency Action Level (EAL) scheme, for NRG review and approval pursuant to 10 CFR 50.54(q)(4) and 10 CFR 50, Appendix E, Section IV.8 .2. The proposed emergency plan will be based on the exemptions requested herein.

Exelon requests review and approval of this exemption request by August 30, 2020. Exelon requests that the approved exemptions become effective 488 days following the permanent shutdown of TMl-1 . TMI will provide the permanent shutdown date in the certification required by 10 CFR 50.82(a)(1)(ii) that TMl-1 has been permanently shutdown and defueled. Approval

U.S. Nuclear Regulatory Commission TMI Request for Exemption Docket Nos. 50-289 and 50-320 July 1, 2019 Page 3 of these exemptions by August 30, 2020, will allow TMl-1 adequate time to implement changes to the emergency plan and emergency response organization by the requested effective date.

This letter contains no new regulatory commitments.

In accordance with 10 CFR 50.91 "Notice for public comment; State consultation" paragraph (b), Exelon is notifying the State of Pennsylvania of this request for exemption by transmitting a copy of this letter and its attachments to the designated State Official.

On May 30, 2019, the Commonwealth of Pennsylvania - Department of Environmental Protection Bureau of Radiation Protection (PA-BRP) received the draft proposed changes of the TMI Permanently Defueled Emergency Plan and EAL scheme. On June 24, 2019, the PA-BRP and Pennsylvania Emergency Management Agency (PEMA) met with representatives of TMl-1 and provided comments on the PDEP and associated EALs. An acknowledgement from the Commonwealth of Pennsylvania confirming that they completed their review of the proposed TMI Emergency Plan/EALs and comments were resolved to their satisfaction will be included in attachment to the License Amendment Request for the proposed changes to the TMI Permanently Defueled Emergency Plan (PDEP) and Emergency Action Level scheme.

If you have any questions concerning this submittal, please contact Leslie Holden at (630) 657-2524.

Respectfully, Michael P. Gallagher Vice President, License Renewal & Decommissioning Exelon Generation Company, LLC

Attachment:

1. Request for Exemptions from Portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and 10 CFR Part 50, Appendix E

2. Three Mile Island Nuclear Station Zirconium Fire Analysis for Drained Spent Fuel Pool (Calculation C-1101-202-E410-476, Revision 1) cc: w/Attachment NRC Regional Administrator, Region I NRC Senior Resident Inspector - Three Mile Island Nuclear Station - Unit 1 NRC Project Manager, NRR - Three Mile Island Nuclear Station - Unit 1 NRC Project Manager, NMSS/DUWP/RDB - Three Mile Island - Unit 2 Director, Bureau of Radiation Protection - PA Department of Environmental Resources

U.S. Nuclear Regulatory Commission TMI Request for Exemption Docket Nos. 50-289 and 50-320 July 1, 2019 Page4 bee: w/o Attachment Sr. Vice President - Mid-Atlantic Operations Site Vice President - TM 1-1 Plant Manager-TMl-1 Director, Operations - TMl-1 Director, Training - TMl-1 wl Attachment Vice President - License Renewal and Decommissioning - KSA Vice President - Licensing and Regulatory Affairs Director, Licensing and Regulatory Affairs (East)

Senior Manager, Emergency Preparedness - Cantara Senior Manager, Decommissioning - Cantara Decommissioning - Licensing - Cantara Site Decommissioning Director- TMl-1 Regulatory Assurance Manager - TMl-1 Manager, Licensing and Regulatory Affairs - KSA Exelon Document Control Desk Licensing (Hard Copy)

Commitment Tracking Coordinator - East Records Management - KSA R. R. Brady - TMl-1 M. D. Fitzwater- TMl-1 E. M. Carreras - TMl-1 C. W. Smith - TMl-1 D. H. Walker - Emergency Preparedness - KSA T. J. Barton - Emergency Preparedness - KSA M. J. Casey, FENOC Fleet Project Management, GPU TMl-2 Project Manager

ATTACHMENT 1 THREE MILE ISLAND NUCLEAR STATION REQUEST FOR EXEMPTIONS FROM PORTIONS OF 10 CFR 50.47(b), 10 CFR 50.47(c)(2)

AND 10 CFR PART 50, APPENDIX E

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 1 of 58 10 CFR Part 50, Appendix E TABLE OF CONTENT ........*.................................................................................................. PAGE 1.0 SPECIFIC EXEMPTION REQUEST ..................................................................................... 2

2.0 BACKGROUND

................................................................................................................... 2 3.0 BASIS FOR EXEMPTION REQUEST .................................................................................. 3 4.0 EXEMPTIONS TO EMERGENCY PLAN REQUIREMENTS DEFINED BY 10 CFR 50.47 AND 10 CFR PART 50, APPENDIX E .................................................................................. 4

5.0 TECHNICAL EVALUATION

................................................................................................36 5.1 Accident Analysis Overview ............................................................................................... 36 5.2 Consequences of Design Basis Events .............................................................................39 5.3 Hottest Fuel Assembly Adiabatic Heat Up (Zirconium Fire) .............................................. .40 5.4 Consequences of Beyond Design Basis Events .......................... ......................................40 5.5 Comparison to NUREG-1738 Industry Decommissioning Commitments and Staff Decommissioning Assumptions ... .... .. ...... .. .... ... ............ .. ...... ....... ......................... ... ........ ..41 5.6 Consequences of a Beyond-Design Basis Earthquake ..................................................... .42

6.0 CONCLUSION

.....................................................................................................................43 7.0 JUSTIFICATION FOR EXEMPTIONS AND SPECIAL CIRCUMSTANCES ........................ 52 7 .1 Exemptions .......................................................................................................................52 7 .2 Special Circumstances ........................................................................... ........................... 53 8.0 PRECEDENT ......................................................................................................................55 9.0 ENVIRONMENTAL ASSESSMENT ....................................................................................55

10.0 REFERENCES

.................................................................:..................................................57

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 2 of 58 10 CFR Part 50, Appendix E 1.0 SPECIFIC EXEMPTION REQUEST Pursuant to 10 CFR 50.12 "Specific exemptions," Exelon Generating Company, LLC (Exelon) requests exemptions from the following for Three Mile Island Nuclear Station:

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 for nuclear power plants; and

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

The Emergency Plan encompasses both Three Mile Island (TMI), Unit 1 (TMl-1), and TMI, Unit 2 (TMl-2). Exelon maintains the emergency planning responsibilities for TMl-2, which is owned by First Energy Corporation, through a service agreement. This exemption request does not impact Exelon's ability to maintain the service agreement.

The requested exemptions would allow Exelon to reduce emergency planning requirements and subsequently revise the TMI Emergency Plan to reflect the permanently defueled condition of the station. The current 10 CFR Part 50 regulatory requirements for emergency planning (developed for operating reactors) ensure safety at TMI. However, once the station is permanently shut down and defueled, and a sufficient decay of the spent fuel has occurred in a state of decommissioning, some of these requirements exceed what is necessary to protect the health and safety of the public.

The requested exemptions and justification for each are based on and consistent with Interim Staff Guidance NSIR/DPR-ISG-02, Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants, which was issued May 11, 2015 (Reference 1).

2.0 BACKGROUND

Three Mile Island Nuclear Station is located in an area of low population density about 12 miles southeast of Harrisburg, Pennsylvania. The area is in Londonderry Township, Dauphin County, about 2.5 miles from the southern tip of Dauphin County, where the county is coterminous with York and Lancaster Counties. The TMI site is part of an 814-acre tract consisting of Three Mile Island and several adjacent islands, which were purchased by a predecessor. The island, which is situated about 900 feet from the east bank and approximately one mile from the west bank of the Susquehanna River, is elongated parallel to the flow of the river with its longest axis oriented approximately due north and south. The north and south ends of the island have access bridges, which connect the island to State Highway Route 441. The north access bridge is used daily. Route 441 is a two-lane highway, which runs parallel to TMI on the east bank of the Susquehanna River and is more than 2,000 feet from the TMI reactors at the closest point. The exclusion area for TMI is a 2,000-foot radius, and for the purposes of Emergency Planning, the exclusion area and the site boundary are considered the same.

Section 6, "Safety Analysis," of the TMl-1 Defueled Safety Analysis Report (DSAR) describes the design basis accident (OBA) scenarios that are applicable to TMl-1. After the reactor is defueled, the spent fuel will be stored in the Spent Fuel Pool (SFP) located in the Fuel Handling Building.

While spent fuel is stored in the SFP, the remaining accident is the Fuel Handling Accident (FHA) that takes place in the SFP.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 3 of 58 10 CFR Part 50, Appendix E The analyses of the potential radiological impact of accidents while the plant is in a permanently defueled condition indicate that no design basis accident or reasonably conceivable beyond design basis accident will be expected to result in radioactive releases that exceed U.S. Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs) (Reference 2) beyond the site boundary. Exelon will maintain the version of the EPA PAGs as specified in the current and proposed TMI Emergency Plan.

TMl-2 has a possession only license and is currently maintained in accordance with the NRC approved SAFSTOR condition (method in which a nuclear facility is placed and maintained in a condition that allows it to be safely stored and subsequently decontaminated) known as Post-Defueling Monitored Storage (PDMS). All fuel assemblies have been removed from the TMl-2 reactor and spent fuel pool.

By letter dated June 20, 2017 (Reference 3), pursuant to 10 CFR 50.82(a)(1 )(i), Exelon submitted a certification to the NRC indicating its intention to permanently cease power operations at TMl-1 on or about September 30, 2019. Once fuel has been permanently removed from the reactor vessel, Exelon 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 TMl-1 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).

By letter dated April 18, 2019 (Reference 4), the NRC issued the Post-Shutdown Emergency Plan (PSEP) which approved changes to the TMI site emergency plan (SEP) to support the planned permanent cessation of operations and permanent defueling of the TMl-1 reactor. The PSEP revised the TMI SEP emergency response organization (ERO) on-shift and augmented staffing, to be commensurate with the reduced spectrum of credible accidents for a permanently shutdown and defueled nuclear power reactor facility. The PSEP maintains effectiveness of the TMI SEP in accordance with 10 CFR 50.47 and 10 CFR 50, Appendix E.

Pursuant to 10 CFR 50.82(a)(4)(i), TMl-1 submitted a Post-Shutdown Decommissioning Activities Report (PSDAR) (Reference 5), which identified SAFSTOR as TMl-1's selected method of decommissioning. With the reactor permanently defueled, the reactor vessel assembly and supporting structures and systems will no longer be in operation and will have no function related to the safe storage and management of irradiated fuel in the SFP. The irradiated fuel will be stored in the SFP and later in the Independent Spent Fuel Storage Installation (ISFSI) (when built) until it is shipped offsite in accordance with the schedules described in the PSDAR and Spent Fuel Management Plan (Reference 6).

3.0 BASIS FOR EXEMPTION REQUEST In order to allow a reduction in emergency planning requirements commensurate with the hazards associated with TMl'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. Exelon has performed an analysis indicating that 488 days after permanent cessation of power operations at TMl-1, a minimum of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is available before fuel cladding temperature reaches 900°C with a complete loss of SFP water inventory with no heat loss (adiabatic heat up). After the 488-day period, there is sufficient time within the 1O hours described in the supporting analysis to mitigate events that could lead to a zirconium cladding fire (herein referred to as the Zirc-Fire Window) (Reference 7). This analysis is contained in Attachment 2. Considering a shutdown date of September 30, 2019, 488 days following

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 4 of 58 10 CFR Part 50, Appendix E permanent cessation of power operations would occur January 30, 2021. Exelon plans to submit a permanently defueled emergency plan (PDEP) by July 1, 2019, 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.

Based on the analyses detailed in Section 5.0, below, Exelon has concluded that the portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR Part 50, Appendix E identified in Tables 1 and 2 will not be necessary to protect the health and safety of the public when TMl-1 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 an appropriate response in the event of an emergency at TMI.

The proposed emergency plan will be based on the exemptions requested herein. Exelon requests approval of these exemption requests by August 30, 2020 with an effective date of meeting the Zirc-Fire Window at 488 days after shutdown, which is expected to be about January 30, 2021. Approval of these exemptions by the requested date will enable Exelon adequate time to implement changes to the emergency preparedness program and emergency response organization.

4.0 EXEMPTIONS TO EMERGENCY PLAN REQUIREMENTS DEFINED BY 10 CFR 50.47AND10 CFR PART 50, APPENDIX E Exelon 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 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 for TMl-1. The specific portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E from which exemptions are being requested are identified using bold strikethreugh 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 bold strikethreugh text (i.e., those portions for which exemption is not being requested) , will remain applicable to TMI. Details related to specific exemption requests are provided in the Basis for Exemption column.

The requested exemptions and justification for each are based on, and consistent with NSIR/DPR-ISG-02 (Reference 1).

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 5 of 58 10 CFR Part 50, Appendix E TABLE 1 EXEMPTIONS REQUESTED FROM 10 CFR 50.47(b) AND (c)(2)

Bold striketlueuah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR 60.47 Emergency Plans Basis for Exemption 1 10 CFR 50.47(b) The onsite and, eXGept as provided in In the Statement of Considerations (SOCs) for the final rule for EP requirements paragraph (d) ef this seGtien, effsite emergency response for independent spent fuel storage installations (ISFSls) and for monitored plans for nuclear power reactors must meet the following retrievable storage (MRS) facilities (60 FR 32430; June 22, 1995) (Reference 8),

standards: the Commission responded to comments concerning offsite emergency planning for ISFSls or MRS and concluded that, "the offsite consequences of potential accidents at an ISFSI or an MRS would not warrant establishing Emergency Planning Zones (EPZs)."

As discussed in ISG-02 (Reference 1), 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 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 ISFSI. Exemptions from offsite EP requirements have previously been approved when the site-specific analyses show that in a partial drain-down event, at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is available from the time when cooling of the spent fuel is not effective until the hottest fuel assembly reaches the zirconium ignition temperature of 900 degrees Celsius (°C). The technical basis that underlies the approval of the exemption request is based partly on the analysis of a time period that spent fuel stored in the SFP is unlikely to reach the zirconium ignition temperature in less than 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . This time period is based on a heat up calculation which uses several simplifying assumptions.

Some of these assumptions are conservative (adiabatic conditions), while others are non-conservative (no oxidation below 900°C). Weighing the conservatisms and non-conservatisms, the staff judges that this calculation reasonably represents conditions which may occur in the event of an SFP accident.

The NRC staff concluded that if 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months were available to initiate mitigative actions, or if needed, offsite protective actions using Comprehensive Emergency

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 6 of 58 10 CFR Part 50, Appendix E TABLE 1 EXEMPTIONS REQUESTED FROM 10 CFR 50.47(b) AND (c)(2)

Bold striketlueuah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR 60.47 Emergency Plans Basis for Exemption Management Plan (CEMP), formal offsite radiological emergency plans would not be necessary for a permanently defueled nuclear power reactor licensee.

As supported by the licensee's SFP analysis, the NRC staff considers an exemption from the requirements for formal offsite radiological emergency plans is justified for a zirconium fire scenario considering the low likelihood of this event together with time available to take mitigative or protective actions between the initiating event and before the onset of a postulated fire.

TMl-1 has an analysis (Reference 9) that demonstrates that 365 days after permanent shutdown, the radiological consequences of the analyzed design basis accident (DBA) 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). An additional analysis (Reference 7) also shows that 488 days after shutdown for an unlikely event of a beyond-OBA where the hottest fuel assembly adiabatic heat up occurs, 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months are available to initiate mitigative or if needed, offsite protective actions, using a CEMP from the time the fuel is uncovered until it reaches the auto-ignition temperature of 900°C.

TMl-1 maintains several strategies implemented by procedures for mitigating the loss of SFP water inventory. These mitigative strategies are maintained in accordance with License Condition 2.c.(17) of the TMl-1 Renewed Facility License. These diverse strategies provide defense-in-depth and can be implemented in ample time to provide makeup water or spray to the SFP prior to the onset of zirconium cladding ignition should a very low probability beyond design basis event affect the SFP.

Several means will be available to provide makeup water to the SFP, such as the Fire Service (FS) System and the portable equipment maintained in accordance with Extensive Damage Mitigating Guidelines (EDMGs) (in support of License Condition 2.c.(17)). There are diverse means to provide makeup water to the SFP

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 7 of 58 10 CFR Part 50, Appendix E TABLE 1 EXEMPTIONS REQUESTED FROM 10 CFR 50.47(b) AND (c)(2)

Bold strikethreueh text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR 60.47 Emergency Plans Basis for Exemption with installed FS electrical and diesel driven pumps, as well as EDMG portable diesel pumps. Water sources are from the river and alternate fire service sources.

Three (3) trained on-shift individuals can implement the established procedures to remove debris, route hoses, and establish an operating portable diesel pump to supply makeup water to the SFP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , well within a 10-hour period.

The three (3) on-shift individuals are assigned to perform this task; they do not have other assigned required emergency preparedness (EP) activities during the performance of this task. Direction and selection of these tasks will continue to be directed by the Certified Fuel Handler and Non-Certified Operator.

Training of the on-shift staff will be maintained, and they will implement such strategies and plans to mitigate the consequences of an event involving a catastrophic loss-of-water inventory concurrently from the SFP.

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 EimergenGy Planning Zenes 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) No exemption requested.

4 10 CFR 50.47{b){3) Arrangements for requesting and Discontinuing offsite emergency planning activities and reducing the scope of effectively using assistance resources have been made, onsite emergency planning is acceptable given the significantly reduced offsite arrangements te assemmedate State and lesal staff at consequences when TMl-1 is in the permanently defueled condition. The TMI the lisensee's Eimergensy Operatiens Fasility ha*.<e been

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 8 of 58 10 CFR Part 50, Appendix E TABLE 1 EXEMPTIONS REQUESTED FROM 10 CFR 50.47(b) AND (c)(2)

Bold strikethra1:1eh text identifies the proposed exemption with respect to the regulation . The basis for the exemption explains the scope of the exception.

Item 10 CFR 60.47 Emergency Plans Basis for Exemption made, and other organizations capable of augmenting the emergency plan will continue to maintain arrangements for requesting and using planned response have been identified. assistance resources from offsite support organizations.

Decommissioning power reactors present a low likelihood of any credible accident resulting in a radiological release together with the time available to take mitigative or, if needed, offsite protective actions using a CEMP between the initiating event and before the onset of a postulated fire. As such, an Emergency Operations Facility would not be required . The Control Room or other onsite location can provide for the communication and coordination with offsite organizations for the level of support required.

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

5 10 CFR 50.47(b)(4) A standard emergency classification and Decommissioning power reactors present a low likelihood of any credible accident action level scheme, the basis of which includes facility resulting in a radiological release together with the time available to take mitigative system and effluent parameters, is in use by the nuclear or, if needed, offsite protective actions using a CEMP between the initiating event facility licensee, aRd State aRd laGal respaRse plaRs Gall and before the onset of a postulated fire. As such, formal offsite radiological fer reliaRGe aR iRfermatiaR provided by faGility liGensees emergency response plans are not required.

fer determiRatians af minim1:1m iRitial affsite respaRse TMI will adopt the Permanently Defueled Emergency Action Levels (EALs) measures consistent with those detailed in Appendix C of Nuclear Energy Institute (NEI) 99-01, "Development of Emergency Action Levels for Non-Passive Reactors,"

Revision 6 (Reference 10), endorsed by the NRC in a letter dated March 28, 2013 (Reference 11 ). A site-specific TMl-1 analysis (Reference 7) shows that after the spent fuel has decayed for 488 days, for beyond design basis events where the SFP is drained, and air cooling is not possible, 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is available to take mitigative or, if needed, 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. No offsite protective actions are anticipated to be necessary. Therefore, classification above the Alert level (e.g.,

Site Area Emergency or General Emergency) will no longer be required.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 9of58 10 CFR Part 50, Appendix E TABLE 1 EXEMPTIONS REQUESTED FROM 10 CFR 50.47(b) AND (c)(2)

Bold strikethrouah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR 50.47 Emergency Plans Basis for Exemption I

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 12), after approximately 1 year of spent fuel decay notification, by the licensee, of State and local response time (and as supported by the SFP analysis), the NRC staff considers an organizations and for notification of emergency personnel by exception to the offsite EPA PAG standard is justified for a zirconium fire scenario all organizations; the content of initial and follow up considering the low likelihood of this event together with time available to take messages to response organizations and the publiG has mitigative or protective actions between the initiating event and before the onset been established; and means to provide early notifiGation of a postulated fire. SECY-13-0112, "Consequence Study of a Beyond-Design-and Glear instruGtion to the populaGe within the plume Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water exposure path*.vay EmergenGy Planning :Zone have been Reactor," (Reference 13) provides that depending on the size of the pool liner established. leak, releases could start anywhere from eight hours to several days after the leak starts, assuming that mitigation measures are unsuccessful. If 10 CFR 50.54(hh)(2)-type mitigation measures are successful, releases could only occur during the first several days after the fuel was removed from the reactor. As previously indicated, a TMl-1 analysis shows that after the spent fuel has decayed for 488 days, for beyond design basis events where the SFP is drained, and air cooling is not possible, 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is available to take mitigative or, if needed, 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. Therefore, offsite emergency plans for the populace within the plume exposure pathway Emergency Planning Zone are not necessary for permanently defueled nuclear power plants.

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 emergency personnel and to the publiG.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 10 of 58 10 CFR Part 50, Appendix E TABLE 1 EXEMPTIONS REQUESTED FROM 10 CFR 50.47(b) AND (c)(2)

Bold striketlueuah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item I 10 CFR 60.47 Emergency Plans I Basis for Exemption 8 I 10 CFR 50.47(b)(7) lnfermatien is made available to the I Refer to basis for 10 CFR 50.47(b).

public en a periodic basis en he*.!f they will be notified and what their initial actions should be in an emergency (e.g., listening to a lecal broadcast station and remaining indoors), lnhe principal points of contact with the news media for dissemination of information during an emergency (including the physical location er lecatiens) are established in advance, and procedures for coordinated dissemination of information to the public are established.

9 I 10 CFR 50.47(b)(8) I No exemption requested .

10 I 10 CFR 50.47(b)(9) Adequate methods, systems, and I Refer to basis for 10 CFR 50.47(b) equipment for assessing and monitoring actual or potential effsite consequences of a radiological emergency condition are in use.

11 10 CFR 50.47(b)(10) A range of protective actions has been TMl-1 has developed an analysis indicating that 488 days after permanent developed for the plume exposure path\\*ay EPZ for cessation of power operations, no credible or beyond design basis accident at emergency workers and the public. In developing this TMl-1 will result in radiological releases requiring offsite protective actions. The range ef actions, censideratien has been given to analysis of the potential radiological impact of the postulated accident for TMl-1 in evacuation, sheltering, and, as a supplement te these, a permanently defueled condition indicates that any releases beyond the site the prophylactic use of potassium iodide (Kl), as boundary are limited to small fractions of the EPA PAG exposure levels.

appropriate. Evacuation time estimates ha*Je been In the unlikely event of a SFP accident, the iodine isotopes which contribute to an developed by applicants and licensees. licensees shall offsite dose from an operating reactor accident are not present, so potassium update the e*Jacuatien time estimates en a periediG iodide (Kl) distribution offsite would no longer serve as an effective or necessary basis. Guidelines fer the choice ef protecti*Je actions supplemental protective action.

during an emergeRGy, consistent 'Nitti Federal guidance, are de*Jeleped and in place, and proteGtive aGtiens for Because it is not possible for PAGs to be exceeded at TMl-1 488 days after permanent cessation of power operations, evacuation planning, including

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 11of58 10 CFR Part 50, Appendix E TABLE 1 EXEMPTIONS REQUESTED FROM 10 CFR 50.47(b) AND (c)(2)

Bold striketlueuah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR 60.47 Emergency Plans Basis for Exemption the ingestion exposure pathway EP-Z appropriate te the evacuation time estimates, is not needed since TMl-1 will meet the criteria for an le6ale have been de*1eleped. exemption from offsite emergency preparedness requirements as discussed in the exemption from 10 CFR 50.47(b).

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

12 10 CFR 50.47(b)(11) through (b)(16) No exemption requested.

13 10 CFR 50.47(c)(1) No exemption requested.

14 10 CFR 50.47(c)(2) Generally, the plume exposure TMl-1 has developed an analysis indicating that 488 days after permanent path~*ay 15PZ fer nu6lear power plants shall 6ensist ef an cessation of power operations, no credible or beyond design basis accident at area about 10 miles (16 km) in radius and the ingestion TMl-1 will result in radiological releases requiring offsite protective actions. The pathway 15PZ shall 6ensist ef an area ahewt 50 miles (80 analysis of the potential radiological impact of the postulated accident for TMl-1 in km) in radiws. The exa6t size and 6enfiguratien ef the a permanently defueled condition indicates that any releases beyond the site 15PZs swrrewnding a parti6wlar nw6lear power rea6ter boundary are limited to small fractions of the EPA PAG exposure levels.

shall he determined in relation te le6al emergen6y Refer to basis for 10 CFR 50.47(b)(10).

response needs and 6apahilities as they are affe6ted by sw6h 6enditiens as demography, topography, land 6hara6teristi6s, a66ess rewtes, and jwrisdi6tienal boundaries. The size of the EPZs alSG 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. The plans fer the ingestion pathv<<ay shall fe6us en su6h a6tiens as are appropriate te prete6t the feed ingestion pathway.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 12 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold striketl:u:oual:I text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption 1 IV Content of Emergency Plans Following docketing of the "Certification of Permanent Removal of Fuel from the Reactor Vessel," in accordance with 10 CFR 50.82(a)(1)(i) and (ii), TMl-1 will

1. The applicant's emergency plans shall contain, but not become a permanently shutdown facility with spent fuel stored in the SFP. In the necessarily be limited to, information needed to EP Final Rule (76 FR 72560, Nov. 23, 2011) (Reference 14), the NRC defined demonstrate compliance with the elements set forth below, "hostile action" as, in part, an act directed toward a nuclear power plant or its i.e., organization for coping with radiological emergencies, personnel. This definition is based on the definition of "hostile action" provided in assessment actions, activation of emergency organization, NRC Bulletin 2005-02, "Emergency Preparedness and Response Actions for notification procedures, emergency facilities and Security-Based Events," dated July 18, 2005(Reference15). NRC Bulletin 2005-02 equipment, training, maintaining emergency preparedness, was not applicable to nuclear power reactors that have permanently ceased

[and] recovery, and onsite protective actions during operations and have certified that fuel has been removed from the reactor vessel.

hostile action. In addition, the emergency response plans submitted by an applicant for a nuclear power reactor The NRC excluded non-power reactors from the definition of "hostile action" at the operating license under this Part, or for an early site permit time of the rulemaking because, as defined in 10 CFR 50.2, a non-power reactor is (as applicable) or combined license under 10 CFR Part 52, not considered a nuclear power reactor and a regulatory basis had not been shall contain information needed to demonstrate developed to support the inclusion of non-power reactors (NPR) in the definition of compliance with the standards described in§ 50.47(b), and "hostile action." Similarly, a decommissioning power reactor or ISFSI is not a they will be evaluated against those standards. "nuclear reactor" as defined in the NRC's regulations. A decommissioning power reactor also has a low likelihood of a credible accident resulting in radiological releases requiring offsite protective measures. For all of these reasons, the NRC staff has concluded that a decommissioning power reactor is not a facility that falls within the definition of "hostile action."

Similarly, for security, risk insights can be used to determine which targets are important to protect against sabotage. A level of security commensurate with the consequences of a sabotage event is required and is evaluated on a site-specific basis. The severity of the consequences declines as fuel ages and, thereby, removes over time the underlying concern that a sabotage attack, under the current definition, could cause offsite radiological consequences.

Although, this analysis provides a justification for an exemption to include the definition for a "hostile action" and its related requirements, elements for security-based events would be maintained. The classification of security-based events,

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 13 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold striketlueueh text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption notification of offsite authorities and coordination with offsite agencies under a CEMP would still be required. Other security-related requirements in the EP Final Rule would be exempted such as, on-shift staffing analysis, emergency response organization (ERO) augmentation and alternative facilities, protection of onsite personnel, and challenging drills and exercises due to the reduced radiological risk for a decommissioning power reactor.

The following similarities between TMI and NPRs show that the TMI facility should be treated in a similar fashion as an NPR. Similar to NPRs, TMI will pose lower radiological risks to the public from accidents than do power reactors because: 1)

TMl-1 will be a permanently shutdown facility (with fuel stored in the SFP and ISFSI) and will no longer generate fission products; 2) fuel stored in the TMl-1 SFP will have lower decay heat resulting in lower risk of fission product release in the event of a beyond design basis boil off or drain down event; and 3) no credible or beyond design basis accident at TMl-1 will result in radiological releases requiring offsite protective actions.

2 2. +his nuGlear P9'-¥er reaGtar liGense appliGant shall Refer to basis for 10 CFR 50.47(b)(10) alse pra¥ide an analysis ef the time required te e¥aGuate '**arieus seGters and distanGes '*'lithin the plume e*pesure path ..¥ay l!P~ fer transient and permanent pepulatiens, using the mast reGent U.S.

Census 8ureau data as ef the date the appliGant suhfflits its appliGatien te the NRC.

3 3. NuGlear P9'-¥er reaGter liGensees shall use NRC Refer to basis for 10 CFR 50.47(b)(10) appre¥ed e¥aGwatien time estimates {li+l!s) and updates te the E+Es in the ferfflulatian af preteGti¥e aGtien reGefflfflendatians and shall pre>1ide the l!+l!!s and l!+I! updates te State and leGal ge¥ernmental authorities fer use in de¥eleping affsite prateGti*1e aGtien strategies.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 14 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethreuah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item I 10 CFR PART 50, APPENDIX E, SECTION IV I Basis for Exemption 4 I 4. Within 365 days ef the later ef the date ef the I Refer to basis for 10 CFR 50.47(b)(10) availability ef the mest reGent deGennial Gensus data fFem the U.S. Census 8ureau er DeGembeF 23, 2011, nuGlear power reaGter liGensees shall develop an ETE analysis using this decennial data and submit it under

§ 50.4 te the NRC. These licensees shall submit this ETi analysis te the NRC at least 1BO days befeFe using it te form preteGtive action recemmendatiens and previding it te State and local governmental authorities fer use in developing effsite protective action strategies.

5 I 5. DuFiRg the years between deGennial Gensuses, I Refer to basis for 10 CFR 50.47(b}(10).

nuGleaF power reaGteF liGensees shall estimate EPZ permanent Fesident pepulatien Ghanges enGe a year, but ne later than 365 days frem the date ef the previous estimate, using the mest F&Gent U.S. Census 8uFeau annual Fesident pepulatien estimate and State/leGal gevemment population data, if available. These liGensees shall maintain these estimates se that they aFe available fer NR.C inspeGtien during the period between deGennial Gensuses and shall submit these estimates te the NR.C 'Nith any updated &TE analysis.

6 I 6. If at any time during the deGennial period, the EPZ I Refer to basis for 10 CFR 50.47(b)(10}

permanent resident pepulatien inGreases suGh that it Gauses the longest ETE value fer the 2 mile zene er 5 mile zene, inGluding all affeGted l!mergenGy Response Planning Areas, eFferthe entire 10 mile &PZte inGrease by 25 perGent er 30 minutes, whiGhever is less, frem the nuGleaF pe*.-.i1er reaGter liGensee's GUrFently NR.C appreved er updated ET!!!:, the liGensee shall update the ETE analysis te refleGt the impact of that pepulatien

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 15 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethrouah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 60, APPENDIX E, SECTION IV Basis for Exemption inGrease. lhe liGensee shall submit the updated ~lli analysis to the NRC under§ 60.4 no later than 366 days after the liGensee's determination that the Griteria fer updating the lilli ha>Je been met and at least ~80 days befere using it to form proteGti¥e aGtion reGommendations and pro>Jiding it to State and loGal go>Jernmental authorities fer use in de¥eloping offsite proteGti¥e aGtion strategies.

7 After an applicant for a combined license< ... > No exemption requested.

8 A. Organization No exemption requested.

The organization for coping < .. . >

9 A.1. A description of the normal plant operating Once TMl-1 is permanently shut down and defueled, a decommissioning reactor organization. will not be authorized to operate under 10 CFR 50.82(a). Because the TMl-1 cannot operate the reactor, a "plant operating organization" will no longer be required.

Rather, the facility will be maintained by a defueled on-shift staff.

10 A.2. No exemption requested .

11 A.3. A desGription, by position and tunGtion to be The number of staff at TMl-1 during decommissioning will be small but perfermed, of the liGensee's headquarters personnel commensurate with the need to safely store spent fuel at the facility in a manner who '/Jill be sent to the plant site to augment the onsite that is protective of public health and safety. TMl-1 will have a level of emergency emergenGy organization. response that does not require response by headquarters personnel. The on-shift and emergency response positions will be defined in the Permanently Defueled Emergency Plan (PDEP).

12 A.4. Identification, by position and function to be performed, TMl-1 has developed an analysis indicating that 488 days after permanent of persons within the licensee organization who will be cessation of power operations, no credible or beyond design basis accident at TMI-responsible for making offsite dose projections, and a 1 will result in radiological releases requiring offsite protective actions. TMl-1 will description of how these projections will be made and the maintain the capability to determine if a radiological release is occurring. If a release is occurring, TMl-1 will promptly communicate that information to offsite authorities

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 16 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethroYAh text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption results transmitted to State and local authorities, NRC, and for their consideration. The offsite organizations are responsible for deciding what, other appropriate governmental entities. if any, protective actions should be taken based on a CEMP.

13 A.5. ldeRtifiGatioR, by positioR aRd fuRGtioR to be As indicated by the TMl-1 adiabatic heat up analysis, the time available to initiate perfermed, of other employees of the liGeRsee '>'lith compensatory actions in the event of a loss of SFP cooling or inventory precludes speGial EIYalifiGatieRs fer GopiRg 'llith emergeRGY the need to identify and describe the special qualifications of these individuals in GORditieRs that may arise. Qther peF&oRs '*'**ith speGial the emergency plan. The number of staff at TMl-1 during decommissioning will be EIYalifiGatioRS, SYGh as GORSYltaRts, who are ROt small but commensurate with the need to maintain the facility in a manner that is employees of the liGeRsee aRd *.Nhe may be Galled YpoR protective of public health and safety.

k>F assistaRGe fer emergeRGies shall also be ideRtified.

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

The speGial EIYalifiGatioRs of these perseRs shall be desGribed.

14 A.6. No exemption requested.

15 A.7. 8y JuRe 23, 2014, [l]dentification of, aRd a desGriptieR A decommissioning power reactor has a low likelihood of a credible accident of the assistance expected from, appropriate State, local, resulting in radiological releases requiring offsite protective measures. For this and Federal agencies with responsibilities for coping with reason and those described in the basis for 10 CFR Part 50, Appendix E, Section emergencies, including hostile aGtieR at the site. Fer IV.1, a decommissioning power reactor is not a facility that falls within the definitions pYrposes of this appeRdix, "hostile aGtioR" is defiRed of "hostile action."

a&-an act directed toward a nuclear power plant or its Similarly, for security, risk insights can be used to determine which targets are personnel that includes the use of violent force to destroy important to protect against sabotage. A level of security commensurate with the equipment, take hostages, and/or intimidate the licensee to consequences of a sabotage event is required and is evaluated on a site-specific achieve an end. This includes attack by air, land, or water basis. The severity of the consequences declines as fuel ages, and over time, the using guns, explosives, projectiles, vehicles, or other underlying concern that a sabotage attack could cause offsite radiological devices used to deliver destructive force.

consequences is removed.

Although the analysis described above and in the basis for 10 CFR Part 50, Appendix E, Section IV.1 provides a justification for exempting TMl-1 from "hostile action" related requirements, some EP requirements for security-based events will be maintained. Protective actions are maintained for onsite personnel through the classification of security-based events, notification of offsite authorities, and

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 17 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethrough text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption coordination of offsite response organizations (i.e., local law enforcement, firefighting, medical assistance) onsite under a CEMP concept.

Refer to basis for 10 CFR Part 50, Appendix E, Section IV.1.

16 A.8. ldentifiGation of the State andtor loGal offiGials Offsite emergency measures are limited to support provided by local police, fire responsible for planning for, ordering and Gontrolling departments, and ambulance and hospital services, as appropriate. Because an appropriate proteGti¥e aGtions, iRGIYding e'ilaGuations analysis has been developed indicating that 488 days after permanent cessation of

.\!hen neGessaty. power operations and due to the low probability of design basis accidents or other credible events to exceed the EPA PAGs, protective actions such as evacuation should not be required, but could be implemented at the discretion of offsite authorities using a CEMP.

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

17 A.9. 8y DeGember 24 1 20~2, for RYGlear po>>¥er reaGtor Responsibilities of the on-shift and emergency response personnel will be detailed liGensees, a detailed analysis demonstrating that on in the Permanently Defueled Emergency Plan and implementing procedures and shift personnel assigned emergenGy plan will be regularly tested through drills and exercises, and audited and inspected by implementation fYRGliORS are not assigned Exelon and the NRC. The duties of the on-shift personnel at a decommissioning responsibilities that i.\IOYld pre¥ent the timely reactor facility are not as complicated and diverse as those for an operating power performanGe of their assigned fYnGtions as speGified in reactor.

the emergenGy plan.

In the EP Final Rule (Reference 14), 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 SFP 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

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 18 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold &tFikethFe1:1gh text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 60, APPENDIX E, SECTION IV Basis for Exemption mitigate an emergency without impeding timely performance of emergency plan functions. For all these reasons, it can be concluded that a decommissioning nuclear power plant {NPP) is exempt from the requirement of 10 CFR Part 50, Appendix E, Section IV.A.9.

18 B. Assessment Actions 8.1. The means to be used for determining the magnitude TMI EALs will be developed consistent with the Permanently Defueled EALs of, and for continually assessing the impact of, the release detailed in Appendix C of NEI 99-01, Revision 6 {Reference 10), which the NRC of radioactive materials shall be described, including found to be an acceptable method for development of EALs. TMl-1 will continue to emergency action levels that are to be used as criteria for review EALs with the Commonwealth of Pennsylvania on an annual basis.

determining the need for notification and participation of However, based upon the reduced scope of EALs for the permanently defueled local and State agencies, the Commission, and other facility, the scope of the annual review of EALs is expected to be limited {i.e.,

Federal agencies, and the emergency action levels that are informal mailings, etc.).

to be used for determining when and what type of protective Also, refer to basis for 10 CFR Part 50, Appendix E, Section IV.1 for the justification measures should be considered within and e1:1tside the site from the requirements in Appendix E related to "hostile action."

boundary to protect health and safety. The emergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and effsite monitoring.

8y J1:1ne 20, 20~2. feF RYGleaF peweF FeaGteF liGensees, these aGtien letJels m1:1st inGIYde hostile aGtien that may adveF&ely affeGt the n1:1Glear pe..,*Jer 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.

19 8.2. No exemption requested.

20 C. Activation of Emergency Organization The Permanently Defueled EALs, developed consistent with Appendix C of NEI 99-01, Revision 6 {Reference 10), will be adopted, as previously described. This

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 19 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethroYoh text identifies the proposed exemption with respect to the regulation . The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 60, APPENDIX E, SECTION IV Basis for Exemption I

C.1. The entire spectrum of emergency conditions that scheme eliminates the Site Area Emergency and General Emergency event involve the alerting or activating of progressively larger classifications.

segments of the total emergency organization shall be Additionally, the need to base EALs on containment parameters is no longer described. The communication steps to be taken to alert or appropriate since these parameters do not provide indication of the conditions at a activate emergency personnel under each class of defueled facility and emergency core cooling systems are no longer required. Other emergency shall be described. Emergency action levels indications, such as SFP level or temperature, can be used at sites where there is (based not only on onsite and off&ite radiation monitoring spent fuel in the SFPs. The EAL scheme presented in NEI 99-01, Revision 6 was information but also on readings from a number of sensors endorsed by the NRC in a letter dated March 28, 2013 (Reference 11 ). No offsite that indicate a potential emergency, SYGh as the pressYre protective actions are anticipated to be necessary, since classification above the in sontainment and the response of the Etmergensy Alert (e.g., Site Area Emergency or General Emergency) level is no longer required.

Core Cooling System) for notification of offsite agencies In the event of an accident at a defueled facility that meets the conditions for shall be described. The existence, but not the details, of a relaxation of emergency planning requirements, there will be available time for message authentication scheme shall be noted for such event mitigation, and if necessary, implementation of offsite protective actions using agencies. The emergency classes defined shall include: (1) a comprehensive approach to emergency planning. See the basis for 10 CFR notification of unusual events, (2) alert, (3) site area 50.47(b) detailing the low likelihood of any credible accident resulting in radiological emergensy, and (4) general emergensy. These classes releases requiring offsite protective measures.

are further discussed in NUREG-0654/FEMA-REP-1.

In the Statement of Considerations for the Final Rule for EP requirements for ISFSls and for MRS facilities (60 FR 32430; June 22, 1995) {Reference 8), 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 TMI that meets the criteria for an exemption from the NRC's offsite EP requirements, there will be time available to initiate mitigative actions consistent with plant conditions, and if necessary, for offsite authorities to employ their CEMP to take protective actions.

As stated in NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants" (February 2001) (Reference 16) for

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 20 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold striketluauah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption 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.

Refer to basis for 10 CFR Part 50, Appendix E, Section IV.B.1.

21 C.2. 8y June 20, 2012, nuGlear pa'J.*.ier reaGter Licensees In the Statement of Consideration for the EP Final Rule published in the Federal shall establish and maintain the capability to assess, Register (76 FR 72560) (Reference 14), non-power reactor licensees were not classify, and declare an emergency condition within 15 required to assess, classify and declare an emergency condition within 15 minutes.

minutes after the availability of indications to plant A SFP and an ISFSI are also not nuclear power reactors as defined in the NRC's operators that an emergency action level has been regulations. A decommissioning power reactor has a low likelihood of a credible exceeded and shall promptly declare the emergency accident resulting in radiological releases requiring offsite protective measures. For condition as soon as possible following identification of the these reasons, the staff concludes that a decommissioning power reactor should appropriate emergency classification level. Licensees shall not be required to assess, classify, and declare an emergency condition within 15 not construe these criteria as a grace period to attempt to minutes.

restore plant conditions to avoid declaring an emergency TMI will maintain the capability to assess, classify, and declare an emergency action due to an emergency action level that has been condition. Emergency declaration is required to be made as soon as conditions exceeded. Licensees shall not construe these criteria as warranting classification are present and recognizable, but within 30 minutes after preventing implementation of response actions deemed by the availability of indications to operators that an EAL threshold has been reached.

the licensee to be necessary to protect public health and In the permanently defueled condition, the rapidly developing scenarios associated safety provided that any delay in declaration does not deny with events initiated during reactor power operation are no longer credible. The the State and local authorities the opportunity to implement consequences resulting from the only remaining events (e.g., fuel handling measures necessary to protect the public health and safety.

accident) develop over a significantly longer period. As such, the 15-minute requirement to assess, classify and declare an emergency is unnecessarily restrictive.

See 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 10 CFR Part 50, Appendix E, Section IV.1.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 21 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethrouah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item I 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption 22 I 0. Notification Procedures D.1. Administrative and physical means for notifying local, I Refer to basis for 10 CFR 50.47(b} and 50.47(b)(10).

State, and Federal officials and agencies and aereements reashed 'Nitti these offisials and agensies fer the pr:empt notifisation of the publiG and fer publis evasuation or other protestive measures, should they besome nesessary, shall be described. This description shall include identification of the appr:epriate offisials, by title and agensy, of the State and local government agencies within the EPZ&.

23 I D.2. Provisions shall be dessribed fer yearly I Refer to basis for 10 CFR 50.47(b} and 50.47(b}(10).

dissemination to the publis 'Nithin the plume exposure pathway EP2 of basis emergensy planning infermation, sush as the methods and times required fer publis notifisation and the protestive astions planned if an assident ossurs, general infermation as to the nature and effests of radiation, and a listing of losal broadsast stations that will be used fer dissemination of infermation during an emergensy. Signs or other measures shall also be used to disseminate to any transient population within the plume exposure pathway EP2 appropriate infermation that woyld be helpfyl if an assident OGGYFS.

24 D.3. A licensee shall have the capability to notify responsible TMl-1 proposes to complete emergency notifications within 30 minutes after the State and local governmental agencies 1Nithin 16 minYtes event classification has been made. This timeframe is consistent with the 10 CFR after declaring an emergency. The lisensee shall 50.72(a)(3) notification to the NRC and is appropriate because in the permanently demonstrate that the appropriate go*Jernmental defueled condition, the rapidly developing scenarios associated with events initiated authorities ha*Je the sapabilit)' to make a publis alerting during reactor power operation are no longer credible and there is no need for State and notifisation desision pr:emptly on being informed or local response organizations to implement any protective actions.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 22 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethrcn1nh text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption by the liGensee af an emergenGy Ganditian. Prier ta Because of the geographic location ofTMI, emergency planning and responsibilities initial operation greater Utan 5 peFGent af rated thermal have historically involved coordination with the Commonwealth of Pennsylvania.

power af the first reaGtar at the site, eaGh nuGlear power Decommissioning-related emergency plan submittals for TMI have been discussed reaGtar liGensee shall demonstrate that administrati'le with offsite response organizations since Exelon provided notification that it would permanently cease power operations at TMl-1. These discussions have addressed and pro'liding prompt invs~r~e~~ established far alerting and physiGal means ha" plume exposure path*.\*ay E~o;s ta th~ publiG with the decommissioning process, including the proposed time of 30 minutes to notify the changes to onsite and offsite emergency preparedness throughout the the prompt publiG alert and na.. he ~es1gn abjeGti'le af ta ha'le the Gapabili~* t state after the event classification has been made. Pennsylvania Emergency t1fiGat1on system shall b y ~am_plete the initial I r t " ¥ o essential! e Management officials have been able to review and concur with this proposal. The a e ing and natifiGatian aUh e~asure path*....ay l!PZ '"'i .e pubhG Within the plume State's acknowledgement of their review will be provided with the PDEP submittal.

use af this alerting and n - .t~1R _about 15 minutes. The Also refer to basis for 10 CFR 50.47(b) and 50.47(b)(10).

fr~m immediate alerting at1f~Gat1o_n Gapability 'Nill range

('Nlt~iR 15 minutes af affiG1als are Ratified th t t:: ti nat1fiGatiaR af the publiG

~e t~at State and laGal urgent aGtion) ta the m:re ~-:*:uat1an exists requiring substantial time a"ail b: e y e¥ents where there is a.

gonernmental authorities r net ta aGti'late the

ata emakfar _e th appropriate

. e a Judgment whether sys~~m. The alerting an:ubh~ al~rt and natifiGatiaR add1t1anally inGlude adm" . nat1~Gat1an Gapability shall far a baGkup method af in1st_rat1¥e and physiGal means Gapable_ of being used i:~~*:.~lerting a~d natifiGatian af alertmg and natifiGatia . vent the_ primary method emergenGy ta alert er natif1~ is una'l~llable during an exposure path"'ay EPZJ all er portions af the plum papulaf e

~~al! ha'le the Gapabili~*

ov t I ion. The baGkup method

.Ath1n the plume expaa~ a a ert and notify the publiG n~ed ta meet the 15 mi=:t:ath*N~y I!~, but does Rat primary prompt publiG alert :::*gn ~~eGti'le far the When there is a deG* . nat1f1Gatian system ISIOR ta aGtiuate th e a Iert and'

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 23 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethreYah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption netifiGatieR system, the apprepriate ge¥ernmental aYtherities 'Nill determine whether te a6ti¥ate the entire alert and netifiGatieR system simYltaneeysly er in a graduated or staged manner. The respensihility fer aGtivating SYGh a puhli6 alert and netifiGation system shall remain with the appropriate go¥ernmeRtal authorities 25 D.4. If FEMA has appro¥ed a RUGlear po*Ner reaGtor Refer to basis for 10 CFR Part 50, Appendix E, Section IV.D.3. regarding the alert site's alert aRd RetifiGatioR design report, iRGluding the and notification system requirements.

haGkYp alert and RotifiGation Gapahility, as of DeGemher 23, 2011, then the haGkYp alert and notifiGation Gapahility requirements in SeGtien IV.D.3 must he implemented hy DeGemher 24, 2012. If the alert and netifiGatien desigR report does not inGIYde a haGkYp alert and notifiGation Gapahility or needs revision to ensure adequate haGkup alert aRd notifiGation Gapahility, then a revision of the alert and RotifiGation design report must he suhmitted to FEMA fer review hy Jyne 24, 2013, and the FEMA appro¥ed haGkup alert and RotifiGatioR meaRs must he implemented within 366 days after FEMA approval. Mowever, the total time period to implemeRt a flii!MA appro¥ed haGkup alert aRd RotifiGatioR means must Rot e>rneed June 22 1 2016.

26 I E. Emergency Facilities and Equipment E.1 thru E.7 No exemption requested.

27 an emereeRGV oeeratiens facility from which effective I

I E.8.a.(i) A licensee oRsite teGhRiGal support GeRter and The TMl-1 analysis indicates that within 488 days after shutdown, no design basis accidents or other credible event at TMl-1 will exceed the EPA PAGs. Due to the low probability of design basis accidents or other credible events to exceed the EPA PAGs at the site boundary, the available time for event mitigation at a

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 24of58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethreuah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption direction can be given and effective control can be exercised decommissioning power reactor and, if needed, to implement offsite protective during an emergency; actions using a CEMP, an emergency operations facility (EOF) would not be required to support offsite agency response. Onsite actions may be directed from the Control Room or other 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 TMI emergency plan will continue to maintain arrangements for requesting assistance and using resources from appropriate offsite support organizations.

Refer to basis for 10 CFR 50.47(b)(3).

28 E.8.a.(ii) Fer RUGlear pe>11er reaGter liGeRsees, a liGeRsee NUREG-0696, "Functional Criteria for Emergency Response Facilities," (Reference eRsite eperatieRal support GeRterj 17) 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 permanently shutdown and 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. The Control Room is the single onsite facility that provides support, emergency mitigation, radiation monitoring, and effective control that will be exercised during an emergency.

29 E.8.b. Fer a RU GI ear peti.tJer reaGter liGeRsee's In accordance with paragraph E.8.e., the requirements of paragraph 8.b do not emeFgeRGy eperatieRs faGility required by paragraph apply to the TMl-1 EOF because it was an approved facility prior to December 23, 8,a et this seGtieR, either a faGili~* leGated hetweeR ~O 2011. However, the exemption is requested to clearly reflect that the requirement miles aRd 26 miles et the nuGlear pe*11er reaGter no longer applies to TMl-1 in a permanently shutdown and defueled condition.

site{s), er a primafY faGility leGated less thaR ~o miles Refer to basis for 10 CFR 50.47(b)(3).

Jrem the nuGlear pe>>*.1er reaGter site{s) and a haGkup faGility le Gated het\'JeeR ~ 0 miles and 26 miles et the RUGlear power reaGter site{s). AR emergeRGY eperatieRs faGility may seF¥e mere than ene nuGlear peti.*..ier reaGter site. A liGensee desiring te leGate an emeFgeRGY operations faGility mere than 25 miles

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 25 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethreueh text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption frem a nuGlear pewer reaGter site shall request prier Cemmissien appreval by submitting an appliGatien fer an amendment te its liGense fer an emergenGy eperatiens faGility leGated mere than 26 miles frem a nuGlear pe*Ner reaGter site, previsiens must he made fer leGating NRC and effsite respenders Gieser te the nuGlear pewer reaGter site se that NRC and effsite respenders Gan interaGt faGe te faGe with emergenGy respense persennel entering and leaving the nuGlear pe*Ner reaGter site, Previsiens fer leGating NRC and effsite respenders Gieser te a nuGlear pe*....er reaGter site that is mere than 26 miles frem the emergenGy eperatiens faGility must inGlude the fellewing:

(1) ~r members f

SpaGeState, federal, ef an NRC site team and an d leGal respenders; (2) Additienal spaGe fer GenduGting briefings with emergenGy respense persennel; (3) CemmuniGa

tion "'I

.. "th ether liGensee and offsite faGilities; emergenGy response

~GGess to plant data and radielogiGal informatien:

(5) Assess to soo*.tino eouioment and office r

&UDO 1es:

30 E.8.c. 8y June 20, 2012, fer a nuGlear pewer reaGter Refer to basis for 10 CFR Part 50, Appendix E, Section IV.E.8.a.(i) and 10 CFR liGensee's emergenGy eperatiens faGility required by 50.47(b)(3).

paragraph 8.a ef this seGtion, a faGility having the fellewing Gapahilities:

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 26 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethrough text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 60, APPENDIX E, SECTION IV Basis for Exemption (1) The Gapability fer obtaining and displaying plaRt data and radielegiGal inferFRatien fer eaGh reaGter at a nuGlear power reaGter site and fer eaGh nuGlear power reaGter site that the faGility serves; (2) The Gapability te ana1yze

plant

teGhniGal inferFRatien and pre¥1 e ~eG1n liGen&ee and eff&ite "d t h iGal briefings en CYVttt Genditiens and. p~gne:~~ e:Gh reaGter at a nuGlear response ergan_1zat1en~ h nuGlear peJJ.*er reaGter power reaGter site and ~r eaG site that

-**- faGilitv

U1e --.

------., GAM'AG'.

- -- - and (3) The Gapability te support response te e¥ents eGGurring siFRultaneeusly at mere than one nuGlear power reaGter site if the eFRergenGy operations faGility serves FRere than one site; and 31 I E.8.d. fer nuGlear pewer reaGter liGensees, an alternati¥e faGility (er faGilities) that would be IRefer to basis for 10 CFR Part 50, Appendix E, Section IV.1. regarding "hostile action."

aGGessible e¥en if the site is under threat ef er experienGing hostile aGtien, te funGtien as a staging area fer augmentation ef eFRergenGy response staff and GelleGti¥ely ha¥ing the felle'.¥ing GharaGteristiGs: the Gapability fer GOFRFRuniGatien with the eFRergenGy operations faGility, Gentrel reeFR, and plant seGurity; the Gapability te perferm eff&ite netifiGatiens; and the Gapability fer engineering assessFRent aGti'fities, inGluding daFRage Gentrel teaFR planning and preparation, fer use when ensite eFRergenGy faGilities Gannet be safely aGGessed during hostile aGtien. The requireFRents in this paragraph 8.d must be iFRpleFRented ne later than DeGember 23, 2014, with the

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 27 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethl'Guah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption ex6eptien ef the Gapahility fer staging emergen6y respense ergani:atien persennel at the alternati¥e fa6ility (er fa6ilities) anEI the Gapahility fer GemmuniGatiens with the emergen6y eperatiens fa6ility, Gentrel ream, anEI plant seGurity, whi6h must he ilQAIAmARtAd

.------------- RA latAr thaR hlRA 20 , 2012 32 E8 A li6ensee s hall net he suhje6t. te, __ the Refer to basis for 10 CFR Part 50, Appendix E, Section IV.E.8.b and 10 CFR

. .e. ~ h 8 h ef this seGt1en ~

50.47(b)(3).

requirements ef paragrap f *s faGility appl'9¥eEI as ef existing emergenGy epera ien ce6emher 23, 2011; 33 E.9.a. Provisions for communications with contiguous Refer to basis for 10 CFR 50.47(b) and (b)(10).

State/local governments within the plume expesure TMl-1 will maintain communications with the Commonwealth of Pennsylvania and path*Nay EPZ. Such communication shall be tested the NRC. Existing commercial phone lines will to be used to communicate EP monthly.

notifications to the Commonwealth of Pennsylvania and will continue to be functionally tested monthly.

34 E.9.b No exemption requested

. eng the nuGlear lh~

35 . "en fer 6emmuni6at1ens a:nsite te6hni6al TMl-1 has developed an analysis indicating that 488 days after permanent p,...,.., , so-I -'"*

emergen6~ e~

E.9.c. oFaliens fa*ilily; cessation of power operations, no credible accident at TMl-1 will result in pewer reaGte El the . al State anEI prlR6~~EI radiological releases requiring offsite protective actions; or in the event of beyond suppert Genter, ::6lear fa6ility, the the field design basis accidents, 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is available to take mitigative actions, and if and among lhe*Y eporaliens sonle':'atieno syete1110 needed, implement offsite protective actions using CEMP concept.

le6al emergen Su6h 6emmun1 ment teams.

assess t El annually. Therefore, there is no need for the Technical Support Center (TSC), Emergency shall he tes e Operations Facility (EOF), or field assessment teams. Additionally, there is no need to maintain and test committed provisions for communications with State and local emergency operations centers (EOCs) with these facilities.

An onsite facility will continue to be maintained, from which effective command and control can be maintained during an emergency. Communication with State and

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 28 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethFeWQh text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 60, APPENDIX E, SECTrlON IV Basis for Exemption local EOCs is maintained to coordinate assistance on site if required. Testing will be as described in justification for 10 CFR 50, Appendix E, Section IV.E.9.a Refer to justification for 10 CFR 50.47(b}(3} and 10 CFR Part 50, Appendix E, Section IV.E.8.a.(i}.

36 E.9.d. Provisions for communications by the licensee with The functions of the Control Room, EOF, TSC and OSC are intended to be NRC Headquarters and the appropriate NRC Regional combined into an onsite facility due to the smaller facility staff and the greatly Office Operations Center from the RWGlear pewer FeaGteF reduced required interaction with State and local emergency response facilities. An Gentrel ream, the ensite teGhniGal swppeFt Genter, and onsite facility will continue to be maintained, from which effective direction can be the emergeRGY eperatiens facility. Such communications given and effective control can be exercised during an emergency. TMl-1 will shall be tested monthly. maintain communication with the NRC.

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

37 I F. Training F.1. The program to provide for: (a) The training of viii. The number of staff at TMl-1 during the decommissioning process will be small employees and exercising, by periodic drills, of radiation but commensurate with the need to safely store spent fuel at the facility in a emergency plans to ensure that employees of the licensee manner that is protective of public health and safety. TMl-1 will maintain a level are familiar with their specific emergency response duties, of emergency response that does not require additional response by and (b) The participation in the training and drills by other headquarters personnel. The on-shift and emergency response positions are persons whose assistance may be needed in the event of a defined in the Permanently Defueled Emergency Plan and will be regularly radiation emergency shall be described. This shall include a tested through drills and exercises, audited, and inspected by Exelon and the description of specialized initial training and periodic NRC.

retraining programs to be provided to each of the following Also see the basis for 10 CFR 50.47(b). Therefore, exempting licensee's categories of emergency personnel:

headquarters personnel from training requirements is considered to be

i. Directors and/or coordinators of the plant emergency reasonable.

organization; ii. Personnel responsible for accident assessment, including control room shift personnel;

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 29of58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethreuah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption iii. Radiological monitoring teams; iv. Fire control teams (fire brigades);

v. Repair and damage control teams; vi. First aid and rescue teams; vii. Medical support personnel; Due to the low probability of design basis accidents or other credible events to viii. biGeRsee's headquarters support perseRRel; exceed the EPA PAGs, offsite emergency measures are limited to support provided ix. Security personnel. by local police, fire departments and medical services, as appropriate. Therefore, the term "Civil Defense" is no longer a commonly used term and is no longer In addition, a radiological orientation training program shall applicable as an example in the regulation. Local news media personnel no longer be made available to local services personnel; e.g., local need radiological orientation training since they will not be called upon to support emergency serviceslCWil DefeRse, local law enforcement the formal Joint Information Center.

personnel, leGal Re'.!.i'S media perseRs.

38 F .2. The plan shall describe provisions for the conduct of TMl-1 analyses demonstrate that 488 days after permanent cessation of power emergency preparedness exercises as follows: Exercises operations, no remaining postulated accidents at TMl-1 will result in radiological shall test the adequacy of timing and content of releases requiring offsite protective actions, or in the event of beyond design basis implementing procedures and methods, test emergency accidents, 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is available to take mitigative actions, and if needed, implement equipment and communications networks, test the puhliG offsite protective actions using a CEMP. Therefore, the public alert and notification alert aRd RetifiGatieR system, and ensure that emergency system will not be used, and no testing would be required.

organization personnel are familiar with their duties.

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

39 F.2.a. A full partiGipatieR e:ic:erGise 'A<hiGh tests as muGh Refer to basis for 10 CFR 50.47(b).

af the liGensee, State, and laGal emergenGy plans as is TMl-1 will continue to invite the Commonwealth of Pennsylvania and local support reasaRahly aGhie,..able ,..,..itheut mandatory publiG to participate in the periodic drills and exercises conducted to assess their ability to partiGipatien shall be GenduGted fer eaGh site at whiGh perform responsibilities related to an emergency at TMI, to the extent defined by a P9'.!.i'er reaGtar is leGated. NuGlear p9'A<er reaGter the TMI emergency plan. Because the need for offsite emergency planning is liGensees shall subtRit eJEeFGise sGenaries under § 50.4 relaxed due to the low probability of the postulated accident or other credible events that would be expected to result in an offsite radioactive release that would exceed

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 30 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold striketlueuah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 60, APPENDIX E, SECTION IV Basis for Exemption at least 60 days befere use in a full partiGipatien the EPA PAGs and the available time for event mitigation, no formal offsite exerGise required by this paragraph 2.a. radiological emergency plans will be in place to test.

F.2.a.(i), (ii), and (iii) are not applicable. The intent of submitting exercise scenarios for use by power reactor licensees is to check that licensees utilize different scenarios in order to prevent the preconditioning of responders at power reactors. For defueled sites, there are limited events that could occur and the previously routine progression to General Emergency in power reactor site scenarios is not applicable to a decommissioning site.

Exelon considers TMI to be exempt from 10 CFR Part 50, Appendix E, Section F.2.a.(i)-(iii) because TMI will be exemptfrom the umbrella provision of 10 CFR Part 50, Appendix E, Section IV.F.2.a.

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

subsequent exercise of its onsite emergency plan every 2 The low probability of design basis accidents or other credible events that would years. NuGlear pe*.,.ier reaGter liGensees shall subFRit result in an offsite radioactive release that would exceed the EPA PAGs and the exerGise sGenaries under § 60.4 at least 60 days befere available time for event mitigative actions at TMl-1 during decommissioning render use in an exerGise required by this paragraph 2.b. :fhe the TSC, OSC and EOF unnecessary. The principal functions required by regulation exerGise FRay be inGluded in the full partiGipatien can be performed at an onsite location that does not meet the requirements of the biennial exerGise required by paragraph 2.G. ef this TSC, OSC or EOF.

seGtien. In addition, the licensee shall take actions necessary to ensure that adequate emergency response TMl-1 will continue to conduct biennial exercises and will invite the Commonwealth capabilities are maintained during the interval between of Pennsylvania and local support organizations (firefighting, law enforcement, and biennial exercises by conducting drills, including at least one ambulance/medical services) to participate in periodic drills and exercises to assess drill involving a combination of some of the principal its ability to perform responsibilities related to an emergency at TMI to the extent functional areas of the licensee's onsite emergency defined by the TMI emergency plan.

response capabilities. The principal functional areas of emergency response include activities such as management and coordination of emergency response, accident assessment, event classification, notification of offsite authorities, and assessment of the onsite and eff&ite

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 31 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethFeuAh text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 60, APPENDIX E, SECTION IV Basis for Exemption impact of radiological releases, proteGti..*e aGtioR reGommeRdatioR developmeRt, pFeteGtit1e aGtioR deGisioR makiRg, plaRt system repair and mitigative action implementation. During these drills, activation of all of the licensee's emergency response facilities (TeGhRiGal Support CeRter (TSC), Oper-atiORS Support CeRter (OSC), and the EmergenGy Operations FaGility (EOF))

would not be necessary, licensees would have the opportunity to consider accident management strategies, supervised instruction would be permitted, 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 training objectives.

41 F.2.c. Offsite plans fer eaGh site shall he ex:erGis~d See basis for 10 CFR Part 50, Appendix E, Section IV.1 and 10 CFR Part 50, hieRRially with full partiGipation by eaG~ off~1te Appendix E,Section IV.F.2.a.

authority ha¥iRg a Fele u~der the . rad1olog1Gal respoRse plan. Where the offs1te authority has a role under a radiologiGal response plan fer more than one site it shall fully partiGipate iR one ex:erGise e*.*ery two yea~ and shall, at least, partially. partiGipate i~ other offsite plan ex:erGises iR this period, If l\..*o d1~reRt liGeRsees eaGh have liGensed faGilities loGated either (1) CoRdUGt aR e:HFGise hieRRiall'I of its ORSite emeFgeRcy plan; (2) Participate euadrenniall'I in an offsite hienRial

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b}, 50.47(c)(2) and Page 32 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50 1 APPENDIX E Bold &triketluouah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 60, APPENDIX E, SECTION IV Basis for Exemption full er partial partiGipatien exerGise; (3) CenduGt emergenGy preparedness aGtivities and interaGtiens in the years between its partiGipatien in the offsite full er partial partiGipatien exerGise with effsite authorities, te test and maintain interfaGe ameng the affeGted State and leGal authorities and the liGensee. Co loGated liGensees shall alse partiGipate in emergenGy preparedness aGtivities and interaGtien

'Nith o~ite authorities for the peried bew.-een exerG1ses; (4) ConduGt a hestile aGtien exerGise ef its rmsit&

emeroenGv olan in eaGt:i exerGise GVGle: and (5) PartiGipate in an e~ite biennial full er paRial partiGipatien hostile aGtien exerGise in alternating exerGise GyGles.

42 I F.2.d. iaGh State with responsibility for nuGlear power I Refer to basis for 10 CFR 50.47(b)(10).

reaGtor emergenGy preparedness should fully partiGipate in the ingestion path*Nay portion of exeFGises at least enGe e*Jery exerGise GyGle. In States with mere than ene nuGlear pe*Ner reaGter plume exposure pathway iPZ, the State sheuld retate this partiGipatien frem site te site. iaGh State with responsibility for nuGlear pev.*er reaGter ernergenGy preparedness sheuld fully partiGipate in a hestile aGtien exerGise at least enGe every GyGle and should fully partiGipate in ene hestile aGtien exerGise by DeGember 31, 2015. States with more than ene nuGlear pewer

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 33 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold &trikethreuah text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 60, APPENDIX E, SECTION IV Basis for Exemption reaGter plume exposure path1May EPZ sheuld relate this partiGipatien frem site te site.

43 F.2.e. Licensees shall enable any State or local Government Refer to basis for 10 CFR 50.47(b)(10).

leGated 'Nithin the plume exposure pathway EPZ to participate in the licensee's drills when requested by such State or local Government.

44 F.2.f. Remedial exercises will be required if the emergency The Federal Emergency Management Agency (FEMA) is responsible for the plan is not satisfactorily tested during the biennial exercise, evaluation of an offsite response exercise. No action is expected from State or local such that NRC,.-iff Gensultatien '.r.'ith FEM.A, cannot (1) find government organizations in response to an event at a decommissioning site other reasonable assurance that adequate protective measures than firefighting, law enforcement, and ambulance/medical services.

can and will be taken in the event of a radiological Memoranda of understanding will continue to be in place for those services. Offsite emergency or (2) determine that the Emergency Response response organizations will continue to take actions to protect the health and safety Organization (ERO) has maintained key skills specific to of the public as they would at any other industrial site.

emergency response. +he e:ic:tent ef State and leGal partiGipatien in remedial exerGises must he suffiGient te she'IJ that apprepriate GerreGti¥e measures ha1.ie been taken regarding the elements ef the plan net preperly tested in the pre¥ieus exerGises.

45 F.2.g and F.2.h No exemption requested.

46 F.2.i. Licensees shall use drill and exercise scenarios that At TMl-1 there will be limited events that could result in radioactive releases that provide reasonable assurance that anticipatory responses exceed the EPA PAGs and the previously routine progression to General will not result from preconditioning of participants. Su&h Emergency in power reactor site scenarios will not be applicable. Therefore, TMl-1 sGenaries fer nuGlear pewer reaGter liGensees must is not expected to demonstrate response to a wide spectrum of events.

inGlude a w<<ide speGtrum ef radielegiGal releases and Also refer to basis for 10 CFR Part 50, Appendix E, Section IV.1 regarding "hostile e¥ents, iAGluding hostile aGtien. Exercise and drill action."

scenarios as appropriate must emphasize coordination among onsite and offsite response organizations.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 34of58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold strikethmuAh text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item I 10 CFR PART 60, APPENDIX E, SECTION IV Basis for Exemption 47 I F.2.j. The exerGises GeRduGted uRder paragraph 2 ef Refer to basis for 10 CFR Part 50, Appendix E, Section IV.F .2.

this seGtien by nYGlear pewer reaGter liGensees Fnust Also refer to basis for 10 CFR Part 50, Appendix E, Section IV.1 regarding "hostile provide the eppertunity fer the ~RO te demenstrate action" and 10 CFR 50.47(b)(5) regarding§ 50.54(hh)(2).

pmfiGieRGY iR the key skills neGessary te implement the prinGipal funGtienal areas ef emergeRGY respense ideRtified in paragraph 2.b ef this seGtien.

EaGh exeFGise must pmvide the eppertunity fer the ERO te demonstrate key skills speGifiG te emergeRGY respeRse duties in the Gentrel reem, TSC, OSC, EOf, aRd jeint infermatien Genter* .Additionally, iR eaGh eight GaleRdar year exeFGise GyGle, RUGlear pewer reaGter liGeRsees shall vary the GenteRt ef sGeRaries duriRg exerGises GeRduGted under paragraph 2 ef this seGtieR te pmvide the eppertuRity fer the ERO te demeRstrate pmfiGieRGy iR the key skills neGessary te respeRd te the felle*.viRg sGenarie elemeRts: hostile aGtien direGted at the plant site, ne radielegiGal release er aR uRplaRned miRimal radielegiGal release that dees Ret require publiG preteGtive aGtieRs, aR iRitial GlassifiGatien ef er rapid esGalatien te a Site Area EmergeRGy er General EmergeRGy, implementatien ef strategies, preGedures, and guidanGe develeped uRder

§ 60.54(hh)(2), aRd integratieR ef effsite reseurGes \'Jith ensite justifiGatieR. The liGeRsee shall maiRtaiR a reGerd ef exerGises GenduGted duriRg eaGh eight year exerGise GyGle that deGumeRts the GeRtent ef sGeRaries used te Gemply with the requiremeRts efthis paragraph. EaGh liGeRsee shall GenduGt a hostile aGtien exerGise fer eaGh ef its sites ne later than DeGember 31, 2015.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 35 of 58 10 CFR Part 50, Appendix E TABLE 2 EXEMPTIONS REQUESTED FROM 10 CFR PART 50, APPENDIX E Bold striketl:u:euAh text identifies the proposed exemption with respect to the regulation. The basis for the exemption explains the scope of the exception.

Item 10 CFR PART 50, APPENDIX E, SECTION IV Basis for Exemption The first eight year e>EerGi&e GyGle fer a site 'Nill hegin in the Galendar year in *.1:hi6h the first hestile aGtien exerGise is GenduGted. fer a site liGensed under Part 62, the first eight year exerGise GyGle hegins in the Galendar year ef the initial e>EeFGise required by SeGtien IV.f.2,a.

48 G. Maintaining Emergency Preparedness and No exemptions requested.

H. Recovery 49 I. Onsite Protective Actions During Hostile Action Refer to basis for 10 CFR Part 50, Appendix E, Section IV.1.

8y June 20, 2012, fer nuGlear pewer reaGter liGensees, a range ef preteGth:e aGtiens te preteGt ensite persennel during hestile aGtien must be develeped te ensure the Gentinued ahility ef the liGensee te safely shut dewn the reaGter and perferm the funGtiens ef the liGensee's emergenGy plan.

NOTE: Appendix E to 10 CFR Part 50, Section Vl.2 exempts permanently or indefinitely shutdown plants from the requirement to provide hardware to support the Emergency Response Data System (EROS). Therefore, specific exemptions from Appendix E to 10 CFR Part 50, sections Vl.1, 3, 4 and 10 CFR 50.72(a)(4) are not required.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 36 of 58 10 CFR Part 50, Appendix E

5.0 TECHNICAL EVALUATION

5.1 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 TMl-1 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 shutdown and defueled condition of TMl-1 is presented below and are in accordance with NRC ISG-02 (Reference 1).

Section 5.0 of ISG-02 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 drain down of the SFP to the point that cooling is not effective, there is at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months (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 events resulting in 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 a 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 TMl-1 addresses each of these analyses.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 37 of 58 10 CFR Part 50, Appendix E TABLE 3 INTERIM STAFF GUIDANCE-02 COMPARISON Analysis ISG-02 Description Response 1 Applicable design DBAs (i.e., fuel handling accident in As discussed in Section 5.2, the postulated design basis accident that will remain the spent fuel storage facility, waste gas system applicable to TMl-1 and could contribute to dose upon implementation of the release, and cask handling accident if the cask requested exemptions is the fuel handling accident (FHA) in the Fuel Handling handling system is not licensed as single-failure-proof) Building, where the SFP is located. The results of the analysis indicate that the dose (Indicates that any radiological release would not at the EAB would not exceed the EPA PAGs 365 days after permanent cessation of exceed the limits of EPA PAGs at EAB); power operations (Reference 9). Exelon will maintain the version of the EPA PAGs as specified in the current and proposed TMI Emergency Plan.

2 Complete loss of SFP water inventory with no heat loss Exelon performed an analysis (Reference 7) that conservatively evaluated the (adiabatic heatup) demonstrating a minimum of 10 length of time (in hours) it takes for uncovered spent fuel assemblies in the SFP to hours is available before any fuel cladding temperature reach the temperature at which the zirconium cladding would fail. The analysis reaches 900 degrees Celsius from the time all cooling concluded that a decay time of 488 days after permanent cessation of power is lost (Demonstrates sufficient time to mitigate events operations is the period that the hottest fuel assembly would reach 900°C in that could lead to a zirconium cladding fire); 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after the assemblies have been uncovered.

This analysis is described in Section 5.3 and is included in Attachment 2.

3 Loss of SFP water inventory resulting in radiation TMl-1 performed an analysis (Reference 18) to determine the radiological impact of exposure at the EAB and control room; (Indicates that a complete loss of SFP water. It was determined that the gamma radiation dose rate any release is less than EPA PAGs at EAB); at the EAB and the Control Room would be less that regulatory defined limits at 488 days after shutdown.

This analysis is described in Section 5.4.

4 Considering the site-specific seismic hazard, either an TMl-1 conducted a seismic evaluation in response to a NRG request for information evaluation demonstrating a high confidence of a low- pursuant to 10 CFR 50.54(f) regarding Recommendation 2.1 of the Near-Term Task probability (less than 1 x 10-5 per year) of seismic Force (NTTF) Review of Insights from the Fukushima Dai-ichi Accident (Reference failure of the spent fuel storage pool structure or an 19). The seismic evaluation included all structures including the SFP, and was analysis demonstrating the fuel has decayed prepared and submitted for NRG review.

sufficiently that natural air flow in a completely drained The Exelon submittal (Reference 20) documents the seismic evaluation in pool would maintain peak cladding temperature below conformance with NTTF Recommendation 2.1 including the high-confidence-of-low-565 degrees Celsius (the point of incipient cladding probability-of-failure (HCLPF) values and the 1 x 1o-5 per year hazard level.

damage) (Indicates that any release is less than EPA PAGs at EAB). The NRG Staff review of the NTTF submittal, specifically for the SFP Evaluation associated with the reevaluated seismic hazard implementing NTTF

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 38 of 58 10 CFR Part 50, Appendix E TABLE 3 INTERIM STAFF GUIDANCE-02 COMPARISON Analysis ISG-02 Description Response Recommendation 2.1 is documented in Reference 21 . The NRC staff concluded that the assessment was performed consistent with the NRG-endorsed (Reference 22)

SFP Evaluation Guidance Report (Reference 23) and provided sufficient information, including the SFP integrity evaluation, to meet the SFP Evaluation Guidance (Item 9 in Enclosure 1 of the NRC's 50.54(f) letter), thus supporting SDA No. 6 of NUREG-1738.

5 The analyses and conclusions described in NUREG- IDCs and SDAs are addressed in Section 5.5 and Tables 4 and 5.

1738 are predicated on the risk reduction measures identified in the study 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 a determination that The onsite restoration plans for repair of the SFP cooling system and to provide there is sufficient resources and adequately trained makeup water to the SFP are incorporated into TMl-1 procedures.

personnel available on-shift to initiate mitigative actions There are multiple ways to initiate mitigative actions and add makeup water to the within the 10-hour minimum time period that will SFP within the 10-hour minimum time period with or without entry to the SFP floor.

prevent an offsite radiological release that exceeds the EPA PAGs at the EAB. Refer to SDA 2 in Table 5.

7 Verify that mitigation strategies are consistent with that TMl-1 maintains procedures and strategies for the movement of any necessary required by the Permanently Defueled Technical portable equipment that will be relied upon for mitigating the loss of SFP water.

Specifications or by retained license conditions. These mitigative strategies were developed in response to 10 CFR 50.54(hh)(2) and are maintained in accordance with License Condition 2.c.(17) of the TMl-1 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.

Refer to SDA 4 in Table 5.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 39 of 58 10 CFR Part 50, Appendix E 5.2 Consequences of Design Basis Events 5.2.1. TMl-1 As described in the license amendment request for proposed changes to the TMI Technical Specifications reflecting the Permanently Defueled condition (Reference 24), the applicable remaining design basis accidents were (1) a Fuel Handling Accident in the Spent Fuel Pool, (2) a Waste Gas Tank Rupture, and (3) a Cask Drop Accident.

As of the end of Zirc-Fire Window all waste gas generated will have been released and the Waste Gas Tank Rupture will no longer be applicable. As stated in the PSDAR (Reference 5), TMI is constructing an ISFSI to support dry fuel storage until the DOE takes possession of the irradiated fuel.

As part of the ISFSI project the Spent Fuel Handling Building Crane is being replaced/upgraded to a 'single failure proof design, and therefore will no longer require a Cask Drop Analysis.

Therefore, the only design basis accident that remain applicable will be the fuel handling accident.

The FHA is defined as the dropping of a single spent fuel assembly in the SFP during fuel handling activities, such that the entire outer row of fuel rods in the assembly, 56 of 208, suffers mechanical damage to the cladding. This accident is postulated to occur despite the administrative controls and physical limitations imposed on fuel-handling operations. The gap activity in the damaged rods is instantaneously released into the SFP. The release occurs under 23 feet of water, which acts as a filter.

The Post Permanent Shutdown FHA (Reference 9) was evaluated using the methodology described in Regulatory Guide 1.183 (Reference 25). This new analysis did not credit the function of any structure, system, or component (SSC) or active mitigation measures. The analysis credits the decontamination of the 23 feet of water over the fuel assemblies in the SFP (i.e., 99.5% (or a Decontamination Factor (OF) of 200) of the iodine released from the fuel assembly is assumed to remain in the water).

The FHA analysis shows that the dose at the EAB 365 days after shutdown (with no credit for safety systems) is 1.78 x 1Q-4 rem TEDE and 5.95 x 10-13 rem Thyroid (Reference 9). This is less than the EPA PAG of 1 rem TEDE and 5 rem Thyroid, and the accepted 10% EPA PAG for declaration of Site Area Emergency per NEI 99-01, Rev.6 (Reference 10).

5.2.2. TMl-2 The bounding event for TMl-2 is a fire in the Reactor Building (RB) with the RB Purge System in operation. Per the TMl-2 Fire Protection Program Evaluation Report (Reference 26) the dose at the exclusion area boundary is 13.5 mrem expressed as a bone dose. Due to the isotopic mix (e.g., negligible amounts of iodine) and the nature of potential releases (i.e., particulate matter),

a more restrictive basis (i.e., the critical organ) for comparison was selected for reporting dose for TMl-2 fires.

This is also less than the EPA PAGs and the accepted 10% EPA PAG for declaration of Site Area Emergency per NEI 99-01, Rev.6 (Reference 10).

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 40 of 58 10 CFR Part 50, Appendix E 5.3 Hottest Fuel Assembly Adiabatic Heat Up (Zirconium Fire)

The analysis (Reference 7) is provided in Attachment 2 to compare the conditions for the hottest fuel assembly stored in the TMl-1 fuel pool to a criterion proposed in SECY-99-168 "Improving Decommissioning Regulations for Nuclear Power Plants" (Reference 27), applicable to offsite emergency response for the unit in the decommissioning process. This criterion considers the time for the hottest assembly to heat up from 30 °C to 900°C adiabatically. If the heat up time is greater than 1O hours, then offsite emergency preplanning involving the plant is not necessary.

Based on the limiting fuel assembly for decay heat and adiabatic heat-up analysis presented in , at 488 days (approximately 16 months) after permanent cessation of power operations, the time for the hottest fuel assembly to reach 900°C is 1O 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 16), 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 1O hours to evacuate) if fuel and cladding oxidation occurs in air.

Because of the length of time it would take for the adiabatic heat up to occur, there is ample time to respond to any 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.

5.4 Consequences of Beyond Design Basis Events NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," (Reference 28) Supplement 1, Section 4.3.9, identifies that a SFP drain down event is a beyond design basis event. The premise of the required adiabatic heat-up analysis was the rapid drain down that would expose the fuel to air cooling. The requirements of the analysis were to determine the decay time required to limit the heat up to 900°C at 1O hours, which would define the mitigation window and event duration.

The offsite and Control Room radiological impacts of a postulated complete loss of SFP water were assessed in Technical Evaluation 623073, "TMI Spent Fuel Pool Draindown Shine Dose Rate Evaluation, Revision O," (Reference 18). A loss of water shielding above the fuel could increase the offsite radiation levels because of the gamma rays streaming up out of the SFP being scattered back to a receptor at the site boundary. With a decay of 365 days from shutdown the dose rate at the EAB would be 4.04 x 10-1 mrem/hr not crediting the shielding from the Fuel Handling Building (FHB) roof. Crediting the FHB roof structure, the dose rate at the EAB would be 4.6 x 10-10 mrem/hr. The resultant dose rates if taken over the 10-hour accident duration would be less than the EPA PAGs and the Site Area Emergency Fraction provided by NEI 99-01, Rev.

6 (Reference 10).

It should be noted that the EPA PAGs were developed to respond to a mobile airborne plume that could transport and deposit radioactive material over a large area. In contrast, the radiation field formed by gamma scatter from a drained SFP would be stationary rather than moving and would not cause transport or deposition of radioactive materials. The extended period required to exceed the EPA 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.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 41of58 10 CFR Part 50, Appendix E Additionally, the Control Room radiological impacts at 365-days of a postulated complete loss of SFP water determined that the gamma radiation dose rate in the Control Room will be below 0.1 mrem/hr.

5.5 Comparison to NUREG-1738 Industry Decommissioning Commitments and Staff Decommissioning Assumptions Although the limited scope of design and beyond design basis accidents that remain applicable to TMl-1 justify a reduction in the necessary scope of emergency response capabilities, Exelon also evaluated the industry decommissioning commitments (IDCs) and staff decommissioning assumptions (SDAs) contained in NUREG-1738 (Reference 16).

NUREG-1738 contains the results of the NRC staff's evaluation of the potential accident risk in spent fuel pools at decommissioning plants in the United States. As stated therein, 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 29). 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 NUREG-1738 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) even though the consequences from a zirconium fire could be serious.

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 staffs analyses and conclusions apply to decommissioning facilities with SFPs that meet the design and operational characteristics assumed in the risk analysis. These

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 42 of 58 10 CFR Part 50, Appendix E 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. 11 The IDCs and SDAs are listed in Tables 4.1-1 and 4.1-2, respectively, of NUREG-1738. The tables below show how the TMl-1 SFP meets or compares with each of these IDCs (Table 4) and SDAs (Table 5).

5.6 Consequences of a Beyond-Design Basis Earthquake NUREG-1738 (Reference 16) 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 a substantial loss-of-water inventory from the SFP, as the only postulated scenario at a decommissioning plant that could result in a 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 2) at the EAB.

However, the risk associated with zirconium cladding fire events decreases as the spent fuel ages, decay time increases, decay heat decreases, and short-lived radionuclides decay away. As decay time increases, the overall risk of a 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 research conducted for NUREG-1738 and N UREG-2161, "Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor," dated September 2014 (Reference 30), suggest 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, regardless of the type of offsite response (i.e., formal offsite radiological emergency preparedness plan or CEMP).

The purpose of NUREG-2161 (Reference 30) 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 SFPs are robust structures that are likely to withstand severe earthquakes without leaking cooling water and potentially uncovering the spent fuel. The study shows the likelihood of a radiological release from the spent fuel 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 the individual 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 1Cl6/year). For such a radiological release, this study shows public and environmental effects are generally the same or smaller than earlier studies. 11

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 43 of 58 10 CFR Part 50, Appendix E The reference plant for the study (a General Electric Type 4 BWR with a Mark I containment) generated approximately 3500 Megawatt-thermal (MWt) and the SFP contained 2844 fuel assemblies. TMl-1 was licensed to generate 2568 MWt, and the SFP has the capacity to hold 1987 fuel assemblies. The SFP is expected to contain 1666 fuel assemblies following permanent cessation of power operations and transfer of all fuel from the reactor vessel to the SFP. Based on these differences, the risk and the consequences of an event involving the SFP at TMl-1 are lower than those in the NUREG-2161 study.

The final off-load into the spent fuel pool will be constrained to ensure that the requirements of Exelon procedure NF-AP-309, "PWR Special Nuclear Material and Core Component Move Sheet Development." Attachment 1, Section 3 - Thermal Management Guidelines in Support of Permanent Shutdown, are met. The off-loaded fuel assemblies (hot cells) will be arranged so that all four face-adjacent cells will have assemblies that have been discharged for at least 5 years (cold cells). Additionally, two or more hot cells may not take credit for the same cold cell. Storing spent fuel in a such a dispersed pattern in SFP promotes air coolability of the spent fuel in the unlikely event of a loss of water. This ensures that fuel distribution in the SFP will be bounded by that assumed in NUREG-2161.

TMl-1 conducted a seismic evaluation in response to a NRC request for information pursuant to 10 CFR 50.54(f) regarding Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident. The seismic evaluation included all structures including the SFP, and was prepared and submitted for NRC review. The Exelon submittal (Reference 20) documents the seismic evaluation in conformance with NTTF Recommendation 2.1 including the high-confidence-of-low-probability-of-failure (HCLPF) values and the 1 x 10-5 per year hazard level. The NRC staff review of the NTTF submittal, specifically for the SFP Evaluation associated with the reevaluated seismic hazard implementing NTTF Recommendation 2.1 is documented in Reference 21. The NRC staff concluded that the assessment was performed consistent with the NRG-endorsed (Reference 22) SFP Evaluation Guidance Report (Reference 23) and provided sufficient information, including the SFP integrity evaluation, to meet the SFP Evaluation Guidance (Item 9 in Enclosure 1 of the NRC's 50.54(f) letter), thus supporting SDA No. 6 of NUREG-1738.

6.0 CONCLUSION

Exelon has concluded, based on the analysis and actions described above, that the health and safety of the public are protected once TMl-1 is in the permanently defueled condition. Approval of the exemptions requested above would not present an undue risk to the public or prevent appropriate response in the event of an emergency at TMI.

Based on the above, TMl-1 has demonstrated that no credible or beyond design basis 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.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 44 of 58 10 CFR Part 50, Appendix E TABLE 4 INDUSTRY DECOMMISSIONING COMMITMENTS (IDCS)

IDC Industry Commitments Response 1 Cask drop analyses will be performed or Currently TMl-1 has analyzed the Fuel Cask Drop Accident in UFSAR Section 14.2.2.8 (Reference single failure-proof cranes will be in use for 31). A fuel cask drop accident is defined as the dropping of a fuel cask through the maximum drop handling of heavy loads (i.e., phase II of height during transfer operations of a fuel cask onto a rail car. A fuel cask drop into the spent fuel pool NUREG-0612 will be implemented). is prevented by the Technical Specification requirement that the key operated travel interlock system for automatically limiting the travel area of the Fuel Handling Building crane shall be imposed whenever loads in excess of 15 tons are lifted and transported.

As discussed in the PDSAR (Reference 5), as part of the ISFSI project the current Spent Fuel Handling Building Crane will be upgraded (or replaced) to a single-failure proof design to handle the spent fuel casks. Since the Spent Fuel Handling Crane will be single-failure proof, the cask drop event will not be considered credible and a cask drop analysis will no longer be required.

2 Procedures and training of personnel will be TMl-1 procedures are in place to ensure onsite and offsite resources can be brought to bear during in place to ensure that onsite and offsite an event, including:

resources can be brought to bear during an

Abnormal Operating Procedure, OP-TM-AOP-035, "Loss of Spent Fuel Pool Cooling" event.

Abnormal Operating Procedure, OP-TM-AOP-008, "Security ThreaVlntrusion"

Abnormal Operating Procedure OP-TM-AOP-020, "Loss of Station Power"

Abnormal Operating Procedure, OP-TM-AOP-002, "Flood"

Abnormal Operating Procedure, OP-TM-AOP-003, "Earthquake"

Abnormal Operating Procedure, OP-TM-AOP-004, "Tornado/High Winds"

ERO activation in accordance with the TMI Permanently Defueled Emergency Plan, EP-TM-1001.

These procedures are required by NRC Regulations and will be implemented as necessary depending on the type of event.

Once TMl-1 is shut down and defueled, the on-shift plant operators, including Certified Fuel Handlers (CFH), and Non-Certified Operators (NCOs) will continue to be appropriately trained on the various actions needed to provide makeup to the SFP based on a systematic approach to training. Once TMl-1 is no longer operating, 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. The

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 45 of 58 10 CFR Part 50, Appendix E TABLE 4 INDUSTRY DECOMMISSIONING COMMITMENTS (IDCS)

IDC Industry Commitments Response TMl-1 CFH training program was approved by the NRC by letter dated December 29, 2017 (Reference 32).

Emergency Plan drills will be conducted to maintain proficiency in response to a plant event as described in the PDEP.

3 Procedures will be in place to establish TMl-1 maintains procedures to provide guidance for establishing and maintaining communications communication between onsite and offsite between offsite agencies and the onsite ERO during severe weather and seismic events.

organizations during severe weather and The following Abnormal Operating Procedures (AOPs) address severe weather and seismic events seismic events.

actions:

Abnormal Operating Procedure, OP-TM-AOP-002, "Flood"

Abnormal Operating Procedure, OP-TM-AOP-003, "Earthquake"

Abnormal Operating Procedure, OP-TM-AOP-004, 'Tornado/High Winds" The AOPs direct entry into the following procedures:

OP-AA-108-111-1001, "Severe Weather and Natural Disaster Guidelines"

OP-TM-108-111-1001, "TMI Severe Weather and Site Inaccessibility Guidelines" These procedures provide direction for additional actions and communications with onsite and offsite stakeholders if the event does not reach the threshold for entry into the PDEP.

If the severity of the event requires entry into the PDEP, communications with onsite and offsite organizations will be directed by the TMI PDEP and associated procedures.

These procedures are required by NRC Regulations and will be implemented as necessary depending on the type of event. Communications are described in the procedures for onsite and offsite communications, they are not specifically referenced in the existing TMl-1 Emergency Plan and will not be included in the planned Permanently Defueled Emergency Plan (to be submitted for NRC approval). Therefore, it is not necessary for them to be specifically referenced in the Emergency Plan. Equipment requirements are specified_in the pertinent procedures.

4 I An offsite resource plan will be developed TMl-1 has multiple portable pumps and emergency generators that meet Extensive Damage which will include access to portable pumps Mitigation Guidelines (EDMG) requirements. These can be used as required by abnormal procedures.

and emergency power to supplement onsite In addition, offsite resources are available from other Exelon Facilities in the nearby vicinity.

resources. The plan would principally identify organizations or suppliers where offsite

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 46 of 58 10 CFR Part 50, Appendix E TABLE 4 INDUSTRY DECOMMISSIONING COMMITMENTS (IDCS)

IDC Industry Commitments Response resources could be obtained in a timely manner.

5 SFP instrumentation will include readouts Spent Fuel Pool Temperature is monitored on the Plant Process Computer and has a high and alarms in the control room (or where temperature alarm function in the Control Room. There are Low Level alarm functions available in the personnel are stationed) for SFP TMl-1 Control Room.

temperature, water level, and area radiation Additionally, there are two channels of continuous remote indication of the SFP water level indicators levels.

in the 322' elevation Control Tower that have been added for reliable SFP level indication (post-Fukushima).

Radiation channel RM-G-9 located in the Fuel Handling Building provides radiation levels in the spent fuel storage area and is monitored and alarmed in the Control Room.

6 SFP seals that could cause leakage leading The TMl-1 SFP is contained in the Fuel Handling Building and is connected to the Fuel Transfer Canal to fuel uncovery in the event of seal failure via two fuel transfer tubes. There are no seals in the SFP that would be subject to leakage.

shall be self-limiting to leakage or otherwise When not actively performing refueling the SFP is isolated from the Fuel Transfer Canal with two engineered so that drainage cannot occur.

blank flanges on the Fuel Transfer Tubes on the Reactor Building side, and two locked closed gate valves on the SFP side. Failure of the Spent Fuel Pool Cooling pump seals will not cause a total drain-down of SFP and is discussed in more detail in IDCS #7.

7 Procedures or administrative controls to The design of the SFP and its cooling system and connections to the pool are such that the SFP reduce the likelihood of rapid drain down cannot be drained below the level of the top of the stored fuel when in its storage rack.

events will include (1) prohibitions on the use The most serious failure of the Spent Fuel Cooling System would be complete loss of water from both of pumps that lack adequate siphon spent fuel storage pools. To protect against this possibility, the cooling water inlet and outlet protection or (2) controls for pump suction connections to spent fuel pool B all enter slightly below, or at, the normal water level in the pool.

and discharge points. The functionality of anti-siphon devices will be periodically Fuel pool A has a drain connection from the spent fuel cooling system extending downward from 10 verified. feet above the top of fuel stored in this pool (330 foot elevation) to 2 inches above the bottom of the pool. This line has a syphon breaker with a normally locked open valve to prevent water from syphoning from the pool below 330-foot elevation in the highly unlikely event that the line should break outside the pool.

A combination drain/fill line enters the spent fuel cask pit at elevation 332 ft (approximately 12 ft

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 47 of 58 10 CFR Part 50, Appendix E TABLE4 INDUSTRY DECOMMISSIONING COMMITMENTS (IDCS)

IDC Industry Commitments Response above the top of the spent fuel stored in pool B). This line extends down inside the pit to elevation 323 ft 6 inches. There is a syphon breaker on this line with a normally locked open valve to prevent draining the spent fuel cask pit below elevation 332 ft in the unlikely event that the line should break outside the pit. The locked open valve is administratively controlled and periodically verified to locked open by operator rounds.

Therefore, it is concluded that the Spent Fuel Cooling System provides adequate protection against serious depression of the water level in either of the spent fuel pools in the highly unlikely event of the rupture of any of its lines.

8 An onsite restoration plan will be in place to There are multiple ways to add makeup water to the SFP with or without entry to the refuel floor.

provide repair of the SFP cooling systems or OP-TM-AOP-035, "Loss of Fuel Pool Cooling", provides the initial response to the abnormal to provide access for makeup water to the conditions in the Spent Fuel Pool. The following procedures describes the spent fuel makeup SFP. The plan will provide for remote strategies:

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

Makeup from Fire Service (OP-TM-251-901, "High Capacity Fire Service Makeup to Spent Fuel Pool")

Makeup from raw water sources (OP-TM-919-922, "FSG Makeup from Raw Water Sources")

Fuel Pool Makeup from FX-P-2A/B (OP-TM-919-914) "Spent Fuel Pool Makeup Using FX-P-2A or FX-P-2B." This method does not require access to the spent fuel pool refueling floor.

Spent Fuel Pool Spray (OP-TM-251-902, "Spent Fuel Pool Spray")

Spent Fuel Pool Spray from outside the SFP Building (OP-TM-251-904, "Spent Fuel Pool Building (External) Spray"), including using an off-site fire truck.

9 Procedures will be in place to control SFP WC-DC-100, "Decommissioning Work Control Process" dictates the review and approval of work operations that have the potential to rapidly conducted while in decommissioning. This procedure directs performance of integrated risk decrease SFP inventory. These assessment per OP-DC-104, "Decommissioning Integrated Risk Management," that provides for administrative controls may require evaluation of potential operational risk.

additional operations or management Heavy loads are controlled through MA-AA-716-022, "Control of Heavy Loads Program." Fuel moves review, management physical presence for and heavy load moves that could affect the safe handling and storage of nuclear fuel require approval designated operations or administrative by the Shift Manager.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 48 of 58 10 CFR Part 50, Appendix E TABLE 4 INDUSTRY DECOMMISSIONING COMMITMENTS (IDCS)

IDC Industry Commitments Response limitations such as restrictions on heavy load Additionally, the ISFSI transfer equipment will be designed such that there will be no ISFSI related movements. SFP operations that will have the potential to cause a rapid drain down of the SFP.

10 Routine testing of the alternative fuel pool TMl-1 has multiple systems and sources to provide alternate makeup to the fuel pool. There is an makeup system components will be electric-driven fire pump (FS-P-2) and a diesel-driven fire pump (FS-P-3) that can supply makeup performed and administrative controls for water to the SFP via the Fire Service System. The TMl-1 fire protection program provides controls for equipment out of service will be implemented operation with equipment out of service and periodic functional testing.

to provide added assurance that the TMl-1 also has two diesel driven engine emergency makeup pumps capable of taking suction from components would be available, if needed.

the river to satisfy the EDMG requirements. The EDMG equipment provides defense-in-depth and have testing and out of service requirements controlled by their program procedures.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 49 of 58 10 CFR Part 50, Appendix E TABLE 5 STAFF DECOMMISSIONING ASSUMPTIONS (SDAS)

SDA Staff Assumptions Response 1 Licensee's SFP cooling design will be at The TMl-1 design aligns with the intent of this description of the standard system in NUREG-least as capable as that assumed in the risk 1738. The TMl-1 Spent Fuel Pool Cooling System (SF) design has two independent trains of assessment, including instrumentation. spent fuel pool cooling. Each train of spent fuel cooling rejects its heat to the Nuclear Service Licensees will have at least one motor-driven Closed Cooling Water System (NSCCW), which in turn rejects its heat to the Susquehanna River and one diesel-driven fire pump capable of (Ultimate Heat Sink) via the Nuclear River Water System (NR).

delivering inventory to the SFP.

Normal makeup to the SFP to provide for evaporation losses is provided by Reclaimed Water.

To provide makeup to address abnormal loss in the spent fuel pool, there are multiple means available. The primary method would be to use Fire Service (FS) water to provide makeup via hoses to the spent fuel pool. The Fire Service System includes a motor driven fire service pump (FS-P-2) and a diesel driven fire pump (FS-P-3), both take suction from the Susquehanna River.

Each FS pump has the capability to deliver 500 gallons per minute (gpm) of makeup water to the SFP. In addition to the river, the fire service system has a water storage tank (Altitude Tank),

which provides an additional 100,000-gallon water source to the FS system.

2 Walk-downs of SFP systems will be Currently TMl-1 performs a walk-down of SFP systems once per day. Once the reactor is performed at least once per shift by the permanently shutdown, shiftly operator rounds will include spent fuel cooling system operating operators. Procedures will be developed for parameters, availability (status) of EDMG and availability of onsite makeup sources. Additionally, and employed by the operators to provide there are other methods available in the Control Room to alert operators to potential SFP events, guidance on the capability and availability of such as annunciators and level indication.

onsite and offsite inventory makeup sources TMl-1 procedure 1104-6, "Spent Fuel Pool Cooling System," describes the normal operation of and time available to initiate these sources the Spent Fuel Pool Cooling system. OP-TM-AOP-035, "Loss of Fuel Pool Cooling," provides the for various loss of cooling or inventory initial response to the abnormal conditions in the Spent Fuel Pool. This AOP will direct mitigation events.

actions related to restoring SFP cooling and/or makeup water. See response for IDC#B for more details on procedures for SFP mitigation strategies.

The ability to use EDMG strategies to provide makeup from the river using portable pumps have been demonstrated to be capable of being implemented within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The operation and control of the Spent Fuel Pooling Cooling Systems and mitigation of a loss of spent fuel pool cooling will be addressed in the Certified Fuel Handling and Non-Certified Operator training programs.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 50 of 58 10 CFR Part 50, Appendix E TABLE 5 STAFF DECOMMISSIONING ASSUMPTIONS (SDAS)

SDA Staff Assumptions Response 3 Control room instrumentation that monitors TMl-1 design meets the intent of this SDA. Spent Fuel Pool temperature is monitored on the plant SFP temperature and water level will directly process computer and has a high temperature alarm function in the Control Room. There are low measure the parameters involved. Level level alarm functions available in the TMl-1 Control Room.

instrumentation will provide alarms at levels Additionally, there are two channels of continuous remote indication of the SFP water level associated with calling in offsite resources indicators in the 322' Control Tower that have been added for reliable SFP level indication (post-and with declaring a general emergency.

Fukushima).

Radiation channel RM-G-9 located in the Fuel Handling Building provides radiation levels in the spent fuel storage area and is monitored and alarmed in the Control Room. Refer to the TMl-1 responses for IDC 2 and IDC 4 for details associated with calling in offsite resources.

Regarding the declaration of a general emergency, the result of the dose calculations for both the Fuel Handling Accident and the beyond design basis event of a total loss of water inventory in the SFP, do not approach the Protective Action Guideline to support a classification of greater than an Alert. TMl-1 will be employing Permanently Defueled 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 (Reference 10). Consistent with the NEI 99-01 Permanently Defueled EALs scheme, it is expected that station conditions will not have the capacity to reach any threshold requiring the declaration of a site area emergency nor general emergency.

4 Licensee determines that there are no drain The TMl-1 SFP design is consistent with this SDA. See discussion for IDC #s 6 and 7.

paths in the SFP that could lower the pool level (by draining, suction, or pumping) more than 15 feet below the normal pool operating level and that licensee must initiate recovery using offsite sources.

5 Load Drop consequence analyses will be The TMl-1 design is in alignment with this description. See discussion for IDC #1.

performed for facilities with non-single failure-proof systems. The analyses and 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

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 51 of 58 10 CFR Part 50, Appendix E TABLE 5 STAFF DECOMMISSIONING ASSUMPTIONS (SDAS)

SDA Staff Assumptions Response confidence in the facilities ability to withstand a heavy load drop.

6 Each decommissioning plant will TMl-1 conducted a seismic evaluation in response to a NRC request for information pursuant to successfully complete the seismic checklist 10 CFR 50.54(f) regarding Recommendation 2.1 of the NTIF Review of Insights from the provided in Appendix 2B to this study Fukushima Dai-ichi Accident. The seismic evaluation included all structures including the SFP,

[NUREG-1738). If the checklist cannot be and was prepared and submitted for NRC review. The Exelon submittal (Reference 20) successfully completed, the documents the seismic evaluation in conformance with NTIF Recommendation 2.1 including the decommissioning plant will perform a plant high-confidence-of-low-probability-of-failure (HCLPF) values and the 1 x 1o-s per year hazard specific seismic risk assessment of the SFP level. The NRC staff review of the NTIF submittal, specifically for the SFP evaluation associated and demonstrate that SFP seismically with the reevaluated seismic hazard implementing NTIF Recommendation 2.1 is documented in induced structural failure and rapid loss of Reference 21. The NRC staff concluded that the assessment was performed consistent with the inventory is less than the generic bounding NRC-endorsed (Reference 22) SFP Evaluation Guidance Report (Reference 23) and provided estimates provided in this study (<1 x10* 5 per sufficient information, including the SFP integrity evaluation, to meet the SFP Evaluation year including non-seismic events). Guidance (Item 9 in Enclosure 1 of the NRC's 50.54(f) letter), thus supporting SDA No. 6 of NUREG-1738 7 Licensees will maintain a program to provide The TMl-1 Spent Fuel Pool "A" contains high density storage racks that employ neutron absorber surveillance and monitoring of Boraflex in material (Baral and Metamic). There are three coupon trees located in the high-density racks.

high-density spent fuel racks until such time Two are located in the Region II racks containing Baral; one is located in the Region II racks as spent fuel is no longer stored in these containing Metamic. Procedure NF-TM-600-1000, "TMI Spent Fuel Rack Boral/Metamic Coupon high-density racks. Program," defines and tracks a surveillance program to verify the long-term integrity of the neutron absorber material used in high-density Spent Fuel Pool storage racks. This program is a license renewal aging management program commitment that has been maintained after permanent cessation of power operations and is required by License Condition 2.(c).21.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 52 of 60 10 CFR Part 50, Appendix E 7.0 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.

7.1 Exemptions 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 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.

The requested exemptions and justification for each are based on and consistent with Interim Staff Guidance NSIR/DPR-ISG-02, Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants, which was issued May 11, 2015.

As discussed in this request, revised radiological analyses have been developed that show that, 365 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the Environmental Protection Agency (EPA) Protective Action Guides (PAGs) at the exclusion area boundary (EAB). In addition, analyses have been developed for beyond design basis events related to the spent fuel pool (SFP) which show that, 488 days after permanent cessation of power operation, 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.

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 0.1 mRem/hr.

Therefore, offsite emergency response plans will no longer be needed for protection of the public beyond the EAB. Based on the reduced consequences of radiological events possible at the site when it is in the permanently defueled condition, the scope of the onsite emergency preparedness organization and corresponding 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. Since 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.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 53 of 60 10 CFR Part 50, Appendix E 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 TMl-1 's ability to physically secure the site or protect special nuclear material. Physical security measures at TMI are not affected by the requested exemption. Therefore, the proposed exemptions are consistent with the common defense and security.

7 .2 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. Exelon has determined that special circumstances are present as discussed below.

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

A. 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 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.

The standards and requirements in 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV 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 TMl-1 are substantially lower than those at an operating plant. The upper bound of offsite 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.

Revised radiological analyses have been developed that show that, 365 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA PAGs at the EAB (Reference 9). In addition, analyses have been developed for beyond design basis events related to the SFP which show that, 488 days (approximately 16 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 PAGs at the EAB (Reference 7).

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 54 of 60 10 CFR Part 50, Appendix E Therefore, application of all of the standards and requirements in 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV 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 TMI being permitted to reduce the scope of emergency preparedness requirements consistent with placing the facility in the permanently defueled condition, the special circumstances are present as defined in 10 CFR 50.12(a)(2)(ii).

B. 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 50, Appendix E, Section IV is not needed for adequate emergency response capability and is excessive for a permanently 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, such as Omaha Public Power District's (OPPD)

Fort Calhoun Station, Unit 1 (FCS); Entergy Nuclear Operation, lnc.'s (ENO) Vermont Yankee Nuclear Power Station (VY); Southern California Edison Company's San Onofre Nuclear Generating Station (SONGS); Duke Energy Florida, lnc.'s Crystal River Unit 3 Nuclear Generating Station (CR3); and Dominion Energy Kewaunee, lnc.'s Kewaunee Power Station (KPS) have been granted similar exemptions.

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.

C. 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 power plant permanently shutdown 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.

The proposed exemptions would allow TMI to revise the station emergency plan to correspond to the reduced scope of remaining accidents and events. As such, the plan would no longer need to address response actions for events that would no longer be possible. The revised plan would thereby enhance the ability of the emergency response organization to respond to those scenarios that remain credible since 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 Emergency Response Organization (ERO) to correctly classify those events that remain credible. As the proposed exemption will enhance the ability of the organization to respond to credible events, a resultant benefit to the public health and safety is realized .

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b}, 50.47(c}(2} and Page 55of60 10 CFR Part 50, Appendix E Therefore, since the 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.

8.0 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 33),

ENO for VY (Reference 34}; to Southern California Edison Company for SONGS, Units 1, 2, and 3 (Reference 35}; to Duke Energy Florida, Inc. for CR3 (Reference 36}; and to Dominion Energy Kewaunee, Inc. for Kewaunee Power Station (Reference 37). 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, Exelon seeks approval of the enclosed proposed exemption requests.

9.0 ENVIRONMENT AL 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 Exelon 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.

When the exemptions become effective, there will be no credible events that would result in doses to the public beyond the exclusion area boundary that would exceed the Environmental Protection Agency (EPA} Protective Action Guides (PAGs}. The probability of occurrence of previously evaluated accidents is not increased, since most previously analyzed accidents will no longer be able to occur and the probability and consequences of the remaining Fuel Handling Accident (FHA} are unaffected by the proposed exemption.

Therefore, the proposed exemption does not involve a significant increase in the probability

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 56 of 60 10 CFR Part 50, Appendix E 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 exemption does not involve a physical alteration of the plant. No new or different type of equipment will be installed and there are no physical modifications to existing equipment associated with the proposed exemption. Similarly, the proposed exemption will not physically change any SSCs involved in the mitigation of any accidents. Thus, no new initiators or precursors of a new or different kind of accident are created. Furthermore, the proposed exemption does not create the possibility of a new accident as a result of new failure modes associated with any equipment or personnel failures. No changes are being made to parameters within which the 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 exemption does not alter the design basis or any safety limits for the plant. The proposed exemption does not impact station operation or any plant SSC that is relied upon for accident mitigation.

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

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

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

There are no expected changes in the types, characteristics, or quantities of effluents discharged to the environment associated with the proposed exemption. There are no materials or chemicals introduced into the plant that could affect the characteristics or types of effluents released offsite.

In addition, the method of operation of waste processing systems will not be affected by the exemption. The proposed exemption will not result in changes to the design basis requirements of SSCs that function to limit or monitor the release of effluents. All the SSCs associated with limiting the release of effluents will continue to be able to perform their functions. Therefore, the proposed exemption will result in no significant change to the types or significant increase in the amounts of any effluents that may be released offsite.

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

The exemption will result in no expected increases in individual or cumulative occupational radiation exposure on either the workforce or the public. There are no expected changes in normal occupational doses. Likewise, design basis accident dose is not impacted by the proposed exemption.

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 57 of 60 10 CFR Part 50, Appendix E (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 TMl-1 emergency response organization commensurate with the reduction in consequences of radiological events that will be possible at TMl-1 once the facility is in the permanently defueled condition. They also will modify the requirements for emergency planning . Therefore, the exemptions address requirements of an administrative, managerial, or organizational nature.

10.0 REFERENCES

1. NSIR/DPR-ISG-02, Interim Staff Guidance, "Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants," dated May 11, 2015 (ADAMS Accession No. ML14106A057)

2. U.S. Environmental Protection Agency, EPA 400-R-92-001, "Manual of Protective Action Guides and Protective Actions Guidelines for Nuclear Incidents," dated October 1991 (reprinted May 1992)

3. Letter from J. Bradley Fewell (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Certification of Permanent Cessation of Power Operations for Three Mile Island Nuclear Station, Unit 1," dated June 20, 2017 (NRC Accession No. ML17171A151)

4. Letter from U.S. Nuclear Regulatory Commission to Bryan C. Hanson, (Exelon Generation Company, LLC), "Three Mile Island Nuclear Station, Units 1 and 2 - Issuance of Amendment No. 296 for Unit 1 RE: Changes to Emergency Plan for Post-Shutdown and Permanently Defueled Condition (EPID L-2018-LLA-0073}, dated April 18, 2019 (ADAMS Accession No. ML19065A114)

5. Letter from Michael P. Gallagher, (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission - "Three Mile Island Nuclear Station, Unit 1 - Post-Shutdown Decommissioning Activities Report," dated April 5, 2019 (ADAMS Accession No. ML19095A041)

6. Letter from Michael P. Gallagher, (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission - "Spent Fuel Management Plan for Three Mile Island Nuclear Station

- Unit 1," dated April 5, 2019 (ADAMS Accession No. ML19095A009)

7. C-1101-202-E410-476, "DE COM Spent Fuel Pool Thermohydraulic Analysis," Revision 1, dated June 10, 2019

8. Federal Register Notice, Vol. 60, No. 120, (60 FR 32430-32442) "Emergency Planning Licensing Requirements for Independent Spent Fuel Storage Facilities (ISFSI) and Monitored Retrievable Storage Facilities (MRS)," dated June 22 , 1995

9. C-1101-900-E000-088, "Fuel Handling Accident Dose Consequence - Post Permanent Shutdown," Revision 0, dated May 11, 2018

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 58 of 60 10 CFR Part 50, Appendix E

10. Nuclear Energy Institute (NEI) 99-01, Revision 6 "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," dated November 2012 (ADAMS Accession No. ML12326A805).

11. Letter from Mark Thaggard (USNRC) to Susan Perkins-Grew (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 (ADAMS Accession No. ML12346A463)

12. U.S. Nuclear Regulatory Commission, Commission Paper SECY-00-0145, "Integrated Rulemaking Plan for Nuclear Power Plant Decommissioning," dated June 28, 2000 (ADAMS Accession No. ML003721626)

13. U.S. Nuclear Regulatory Commission, Commission Paper SECY-13-0112, "Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor," dated October 9, 2013 (ADAMS Accession No. ML13256A339)

14. Federal Register Notice, Vol. 76, No. 226 (76 FR 72560), Enhancements to Emergency Preparedness Regulations, dated November 23, 2011

15. U.S. Nuclear Regulatory Commission, Bulletin 2005-02, "Emergency Preparedness and Response Actions for Security-Based Events," dated July 18, 2005 (ADAMS Accession No. ML051740058)

16. NUREG-1738, "Technical Study of Spent Fuel Accident Risk at Decommissioning Nuclear Power Plants," dated February 2001 (ADAMS Accession No. ML010430066)

17. U.S. Nuclear Regulatory Commission, NUREG-0696, "Functional Criteria for Emergency Response Facilities," dated February 1981 (ADAMS Accession No. ML051390358)

18. Technical Evaluation 623073, "TMI Spent Fuel Pool Draindown Shine Dose Rate Evaluation, Revision O," dated May 28, 2018

19. Letter from U.S. Nuclear Regulatory Commission to All Power Reactor Licensees, "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, And 9.3, of The Near-Term Task Force Review of Insights from The Fukushima Dai-lchi Accident," dated March 12, 2012 (ADAMS Accession No. ML12073A348)

20. Letter from Mr. James Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Exelon Generation Company, LLC, Seismic Hazard and Screening Report (Central and Eastern United States (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," dated March 31, 2014 (ADAMS Accession No. ML14090A271)

21. Letter from U.S. Nuclear Regulatory Commission to Mr. Bryan C. Hanson (Exelon Generation Company, LLC), "Three Mile Island Nuclear Station, Unit 1 - 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 Dal-ichi Accident (CAC NO. MF3905)," dated August 14, 2015 (ADAMS Accession No. ML15223A215)

22. Letter, Jack R. Davis (USNRC) to Joseph E. Pollock (NEI), "Endorsement of Electric Power Research Institute Report 3002007148, Seismic Evaluation Guidance: Spent Fuel Pool Integrity Evaluation," dated March 17, 2016 (ADAMS Accession No. ML15350A158)

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 59 of 60 10 CFR Part 50, Appendix E

23. EPRI, "Seismic Evaluation Guidance: Spent Fuel Pool Integrity Evaluation," Electric Power Research Institute Technical Update 3002007148, dated February 2016 (ADAMS Accession No. ML16055A021)

24. Letter from Michael P. Gallagher, (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission - "License Amendment Request - Proposed Defueled Technical Specifications and Revised License Condition for Permanently Defueled Conditions," dated July 25, 2018 (ADAMS Accession No. ML18206A545)

25. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,"

dated July 2000 (ADAMS Accession No. ML003716792)

26. 990-3017, "Three Mile Island Unit No. 2 Fire Protection Program Evaluation, Revision 12, dated May 18, 2018

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

28. U.S. Nuclear Regulatory Commission, NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," dated October 2002

29. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated May 2011 (ADAMS Accession No. ML100910006)

30. U.S. Nuclear Regulatory Commission, NUREG-2161, "Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor,"

dated September 2014 (ADAMS Accession No. ML14255A365)

31. Three Mile Island Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 14, "Safety Analysis," Section 14.2.2.8, "Fuel Cask Drop Accident," Revision 24, April 2018

32. Letter from U.S. Nuclear Regulatory Commission to Mr. Bryan C. Hanson (Exelon Generation Company, LLC}, "Three Mile Island Nuclear Station; Unit 1 - Approval of Certified Fuel Handler Training and Retraining Program (CAC NOS. MF9960, EPID L-2017-LLL-0013)" dated December 29, 2017 (ADAMS Accession No. ML17228A729)

33. U.S. Nuclear Regulatory Commission, Omaha Public Power District, Fort Calhoun Station, "Fort Calhoun Station, Unit No. 1 - Exemptions from Certain Emergency Planning Requirements and Related Safety Evaluation (CAC NO. MF9067; EPID L-2016-LLE-0003)," Dated December 11, 2017, (ADAMS Accession No. ML17263B191)

34. Federal Register Notice, Vol. 80, No. 242 (80 FR 78776), Entergy Nuclear Operations, Inc.;

Vermont Yankee Nuclear Power Station, Exemption; issuance, dated December 17, 2015

35. Federal Register Notice, Vol. 80, No. 113 (80 FR 33558), Southern California Edison Company; San Onofre Nuclear Generating Station, Units 1, 2, and 3, and Independent Spent Fuel Storage Installation, Exemption; issuance, dated June 12, 2015

36. Federal Register Notice, Vol. 80, No. 69 (80 FR 19358), Duke Energy Florida, Inc.; Crystal River Unit 3 Nuclear Generating Station, Exemption; issuance, dated April 10, 2015

Request for Exemption from Portions of Attachment 1 10 CFR 50.47(b), 50.47(c)(2) and Page 60 of 60 10 CFR Part 50, Appendix E

37. Federal Register Notice, Vol. 79, No. 214 (79 FR 65715), Dominion Energy Kewaunee, Inc.;

Kewaunee Power Station, Exemption; issuance, dated November 5, 2014

ATTACHMENT 2 THREE MILE ISLAND NUCLEAR STATION THREE MILE ISLAND NUCLEAR STATION ZIRCONIUM FIRE ANALYSIS FOR DRAINED SPENT FUEL POOL (CALCULATION C-1101-202-E410-476, Revision 1)

Des1an. . cover Sh eet A na1ys1s Design Analysis I Last Page No. 66 Analysis No.: C-1101-202-E410-476 Revision: 1 Ma1or 181 Minor 0

Title:

DECOM Spent Fuel Pool TH Analysis EC No.:* 623197 Revision: 0 Station(s): Three Mile Island Component(s):

Unit No.:* 01 N/A Discipline:

TEDM Descrlp. Code/Keyword: Spent Fuel Safety/QA Class: Augmented Quality System Code: 202 Structure: ** N/A CONTROLLED DOCUMENT REFERENCES

Document No.: From/To Document No.: From/To ER-TM-TSC-0016 Rev 7 From Hl-89407 Rev 6 From Is this Design Analysts Safeguards Information? " YesO No 181 If yes, see SY-AA-101-106 Does this Design Analysis contain Unverified Assumptions? '

YesO No 181 If yes, ATl/AR#:

This Design Analysis SUPERCEDES: 11 None in its entirety.

Description of Revision (list changed pages when all pages of original analysis were not changed): " See Revision Summary page.

Dan Sells I /J~j-;I{ c-/~0/11 Preparer:

8111 Mcsorley Punl Nam*

0 , I S.on Name 1,.J:s/17 Dale Method of Review: Detailed Review 181 Alternate Cai~ations (attached) D Testing D Reviewer:

Mara Levy I _; - c 3.-- s/go/Jq

~~ s-x">-'eoi Greg Heasley PnnlName - Dall Review Notes: Independent review 181 Peer review 0 An independent review of the product was performed IAW CC-AA-309, Rev. 11 and CC-AA-309-1001, Rev. 9. The minimum decay time after shut down of TMl-1 Cycle 22 in Attachment 1 were determined to be acceptable given the inputs, assumptions, and methodology used.

1r or ~em ll Analv*u. On.vi External Approver: ,. NIA NIA NIA Prnt Name Sogn Noma - Cale Exelon Reviewer: " NIA NIA NIA Pr1nl Name -- Soan Name Oa1a Independent 3rd Party Review Reqd? " YesO No 181

~~Ch-A- - l>/ll/1'/

Exelon Approver:

Patrick Brady I

- ~~

~~~/,~.-_, ~J - b/1P/Jtj Patrick Bennett

- Pron! Name v 'l.,,, Name . C1ale

Exelon WO DESIGN ANALYSIS SHEET Nuclear

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

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202 I Sheet 2 of 66 Table of Contents Cover Page 1 Table of Contents 2 Revision Summary 3 1.0 Purpose 4 2.0 Results and Conclusions 5 3.0 References and Computer Programs 9 4.0 Assumptions 11 5.0 Design Input 12 6.0 Method of Analysis & Numerical Analysis 16 7.0 Appendices 25 Attachment #1: ORIGEN2 Decay Heat Calculation for Fuel Assembly with Maximum Decay Heat Load Attachment #2: Specific Heat Study

DESIGN ANALYSIS SHEET Exelon""

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

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202 I Sheet 3of66 Revision Summary Revision 1

Added the following to the available mass to heat up

- (7) M5 Spacer Grids

- (1) HMP Spacer Grid

- Upper End Fitting

- Lower End Fitting

Adjusted the lengths of the fuel rods, guide tubes, and instrument tube to be their actual lengths instead of just the fuel stack length.

Added references for fuel rod and instrument tube lengths.

Repeated the Purpose #2 calculation at 1 year after shutdown as a revised input for Purpose #5.

Changed the initial temperature for Purpose #5 to just above the maximum design temperature calculated using the decay heat at 1 year.

Added references for heat capacity/specific heat for CF3 stainless steel (end fittings) and lnconel Alloy 718 (HMP Spacer Grid).

Changed the temperature used to calculate the heat capacities to the midpoint temperature rather than the 500F value and recalculated the heat capacities.

Updated the specific heat study (Attachment 2) with the addition of CF3 stainless steel and lnconel Alloy 718.

Revised the Decay Heat Generation Rate calculation to include CF3 stainless steel and lnconel Alloy 718 components.

Adjusted section and equations numbers as necessary.

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202 I Sheet 4of66

1. Purpose

1. Establish the total heat generation rate in the spent fuel versus time after permanent shutdown in 2019 and all fuel is loaded into the SFP.

2. Determine spent fuel pool water temperature with one or both cooling trains in operation, assuming the reactor has been shut down for 14 days, and NSCCW temperature to the spent fuel coolers is 95°F. This supports DSAR.

3. Determine the "time to boil" (TTB) and "time to Top of Active Fuel" (TTAF) versus time after permanent shutdown in 2019 and all fuel is loaded into the SFP.

4. Determine the spent fuel assembly with the highest heat generation rate, and determine the heat generation rate of that assembly versus time after reactor shutdown.

5. Determine the minimum decay time for the limiting spent fuel assembly (purpose #4) to heat up to 900°C (1652°F/1173 K) in 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after a loss of all cooling. This is required for EPlan exemption in accordance with Reference 3.3.

Background:

1. Purpose #1 is required to perform calculations for Purpose #2 and #3. This result is not required for a purpose outside this analysis.

2. Purpose #2 is required for the Defueled Safety Analysis Report (DSAR) section 3.2. This DSAR section is a revision of UFSAR section 9.4. This revision replaces the maximum SF pool temperatures based on the "bounding" design basis SFP heat load with the temperatures based to the maximum heat with the defueled spent fuel pool heat load. This result will be used to change a design basis parameter in the UFSAR/DSAR. The spent fuel pool temperature will also be calculated at 1 year after shutdown to support Purpose

5.

3. Purpose #3 is required for evaluation of loss of spent fuel pool cooling mitigating strategies, and the required response time. The cause for the loss of spent fuel pool cooling includes fire, aircraft impact, multiple equipment failures, loss of offsite power, etc.

where the spent fuel pool integrity is not affected. This result does not change a design basis parameter.

4. Purpose #4 is required to perform calculations for Purpose #5. This result is not required for a purpose outside this analysis.

5. Purpose #5 is required to support a request for exemption to Eplan requirements. The regulatory basis for this exemption is described in reference 3.3 section 5 item 2. This result does not change a design basis parameter.

Exelon)M DESIGN ANALYSIS SHEET Nuclear

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202 I Sheet 5of66

2. Results and Conclusions 2.1. The total heat generation rate in the spent fuel versus time after permanent shutdown in 2019 and all fuel is loaded into the SFP is shown in Figure 2.1 below. Tabular values of this result are in Appendix 7.1.

Figure 2.1 - Post-Defueled SFP Decay Heat 7 00 G 00 5 00

5

s::::;; 4 00 200 1 00 000 o 200 400 600 800 1000 1200 1400 1600 Time after Shutdown (days)

The spent fuel pool decay heat at 14 days after shutdown is 6.38 MW1t1, and the spent fuel pool decay heat at 1 year after shutdown is 1.57 MWtti (input for purpose #2).

2.2. Maximum SFP Temperature Results 2.2.1 . With one SF cooling train in service at 14 days after shutdown, the maximum SF pool temperature is 179.7°F.

2.2.2. With both SF cooling trains in service at 14 days after shutdown, the maximum SF pool temperature is 137.4°F.

Exelonw DESIGN ANALYSIS SHEET Nuclear

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

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202 I Sheet 6of66 2.2.3. The maximum "expected" SFP temperature at 14 days after shutdown is 122.4°F (input for purpose #3).

2.2.4. With one SF cooling train in service at 1 year after shutdown, the maximum SF pool temperature is 115.9°F.

2.2.5. With both SF cooling trains in service at 1 year after shutdown, the maximum SF pool temperature is 105.4°F (input for purpose #5).

2.2.6. The maximum "expected" SFP temperature at 1 year after shutdown is 90.4°F.

2.3. SFP TTB/TTAF 2.3.1 . The time to boil after permanent shutdown is shown in Figure 2.3.1 below. Tabular values of this result are in Appendix 7.3.

Figure 2.3 1 - Post-Defueled Time to Boil 1 40 Ui' 80 5

5al g

CIJ E 60

~

1600 Time after Shutdown (days)

Exelon . DESIGN ANALYSIS SHEET Nuclear

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

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202 Sheet 7 of66 2.3.2. The time to top of active fuel after permanent shutdown is shown in Figure 2.3.2 below.

Tabular values of this result are in Appendix 7.3.

Figure 2.3.2 - Post-Defueled Time to Top of Active Fuel BO 1600 Time after Shutdown (days)

Exelon)M DESIGN ANALYSIS SHEET Nuclear

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

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202 I Sheet 8ot66 2.4. A best estimate decay heat load for the fuel assembly with the maximum heat load in the TMI SFP following TMl-1 Cycle 22 shutdown has been calculated using the ORIGEN2 computer code. Per Attachment 1 Section 7.3 of this calculation, the decay heat load from this Batch 22A fuel assembly as a function of decay time (in days) after Cycle 22 shutdown is shown in the table below. This decay heat load is appropriate to be used as input to the Beyond Design Basis zirconium fire event for a drained SFP.

Maximum Fuel Assembly Decay Heat (Watts) for Various Decay Times After TMl-1 Cycle 22 Shutdown Decay Time (Days) 365.0D 548.0D 730.00 913.00 1095.0D Decay Heat (W) 6.44E+03 4.70E+03 3.65E+03 2.94E+03 2.44E+03 The following polynomial fit represents this data for decay heat (x) between 2.44 and 6.44 kW.

The polynomial fit is in good agreement with the individual ORIGEN2 data points. A 1% multiplier is applied to the Excel fit for conservatism.

Y (days)= 1.01*[-9.22479E-09*x3 +1.59163E-04*x2 -1.01332E+OO*x + 2.75333E+03]

2.5. The minimum decay time for the limiting spent fuel assembly (Batch 22A) to heat up to 900°C (1652°F/1173K) in 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after a loss of all cooling is 488 days.

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202 I Sheet 9of66

3. References and Computer Programs 3.1. CC-AA-309 "Control of Design Analyses" 3.2. CC-AA-309-1001, "Guidelines for Preparation and Processing of Design Analyses" 3.3. NISR/DPR-ISG-02 "Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants" 3.4. FS1-0030101, Rev. 1, TMl-1Cycle22 Final Fuel Cycle Design 3.5. EX0009886 RE Decay Heat Version 0 3.6. Exelon Nuclear Fuels TMI Fuel Database, "C22database.accdb," 1/03/2018.

3.7. Exelon TODI NF173266, Rev. 0, "TMI End-of-Cycle 21 Data and FIDMS Files," 9/18/2017.

3.8. Exelon TODI NF173357, Rev. 0, "TMl-1 Cycle 22 Control Rod Drop Times and Start-up Data,"

10/23/2017.

3.9. UFSAR Section 9.4 3.10. Hl-89402, Rev. 2, Thermal Hydraulic Analysis of TMl-1 Spent Fuel Pool, dated September 30, 1990.

3.11. Hl-89407 Holtec International, Licensing Report for Pool A Reracking, Rev. 6, dated October 30, 1990 (UFSAR Chapter 9 Reference 10) 3.12. C-1101-202-E270-438, Rev. 1, Core and SFP Time to Boil/Uncover 3.13. 1505-1, Rev. 62, Fuel and Control Component Shuffles 3.14. ER-TM-TSC-0016, Rev. 7, RCS and SFP Heatup and Inventory Boiloff Following Loss of Active Decay Heat Removal 3.15. FS1-0029697, Rev. 1, Task 76 Inputs, Mechanical Design Information for Safety Analysis, and Regulatory Compliance for TMl1-22 3.16. "Chart of the Nu cl ides", Thirteenth Edition. General Electric, 1984.

3.17. FS1-0036046-1.0 Confirmation of Material Properties for TMI 3.18. Kim, Cheong S. "Thermophysical Properties of Stainless Steels." Argonne National Laboratory, 1975.

3.19. Farwick, D. G., and R. N. Johnson. "Thermophysical Properties of Selected Wear-Resistant Alloys." U.S. Dept. of Energy, 1980.

3.20. AREVA Dwg. 9017973-000, MK-B-HTP Fuel Bundle Assembly

Exelon~

DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis 'Rev. No. , System Nos. , Sheet C-1101-202-E410-476 1 202 10 of66 3.21. Framatome Dwg. 1238726, Rev 1, Instrumentation Tube (MONOBLOC) 3.22. Esmaili, H. "Spent Fuel Assembly Heat Up Calculations in Support of Task 2 of User Need NSIR-2015-01 ,"April 2016 (ADAMS Accession No. ML16110A431).

Exelon~

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

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202 I 11 Sheet of 66

4. Assumptions 4.1. Purpose #1 Assumptions 4.1.1. Fuel batch burnups are taken from NEMO calculations documented in Tables 38-39 and Figure 34 of Reference 3.4. NEMO is the design code for TMl-1 core simulations and calculated burnups are representative of actual burnups measured by the core monitoring system. Burnups prior to Cycle 21 are based on core follow calculations that are depleted to actual cycle lengths. Cycle 21 burnups are based on a design depletion to 647 EFPD; differences from the actual Cycle 21 length of 648.8 EFPD (Reference 3.7) are negligible with respect to this decay heat calculation. Cycle 22 burnups are based on 720 EFPD, which is the maximum licensed cycle length for Cycle 22. Higher fuel assembly burnups are conservative for decay heat calculations.

4.1.2. Reference 3.4 lists fuel batch enrichments based on the nominal base enrichment of the batch. Since lower enrichments yield higher decay heat results, actual enrichments based on NRC 741 forms are used in this calculation. This includes segments of the fuel assemblies that have lower than nominal enrichments (i.e., Gad rods and axial blankets).

The actual enrichments were obtained from Reference 3.6.

4.1.3. End-of-Cycle 22 date: 30 September 2019 Basis: This date is conservative and corresponds to the maximum cycle length based on a licensed cycle length of 720 EFPD.

4.2. Purpose #2 Assumptions 4.2.1 . No assumptions were made in this section.

4.3. Purpose #3 Assumptions 4.3.1. The initial SFP temperature is 125°F.

Basis: This value is conservative based on the maximum "expected" SFP temperature of 122.4°F determined in section 6.2.1 of this calculation . Setting the initial temperature at the maximum "expected" pool temperature is conservative because the heat-up time would be increased with a lower starting temperature.

4.3.2. The initial SFP level is assumed to be at the low-level limit of 343'-6".

Basis: This value is conservative based on the normal operating band of the SFP of 344'-

5" to 344'-9" (Reference 3.13 Figure 1) because less water would result in a lower heat-up time.

4.4. Purpose #4 Assumptions See Attachment 2 Section 4.

Exelon WO DESIGN ANALYSIS SHEET Nuclear

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202 I 12Sheet of66 4.5. Purpose #5 Assumptions 4.5.1 . No assumptions were made in this section.

5. Design Input 5.1. Purpose #1 Design Inputs 5.1.1. Batch-average burnups for fuel loaded in Cycle 22 are taken from Tables 38-39 of Reference 3.4.

5.1.2. Actual batch enrichments for fuel loaded in Cycle 22 are taken from Reference 3.6. For each batch, the lowest as-built enrichment of all fuel assemblies in the batch is used to represent the batch.

5.1.3. Inputs for every fuel assembly currently stored in the spent fuel pool (burnup, enrichment, cycles in core, batch number) are taken from the TMI Fuel Database (Reference 3.6).

5.1.4. End-of-Cycle 21 date/time (reactor shutdown): 18 September 2017 at 05:16 (Reference 3.7).

5.1.5. Beginning-of-Cycle 22 date/time (initial criticality): 07 October 2017 at 17:43 (Reference 3.8).

5.2. Purpose #2 Design Inputs 5.2.1. SFP Heat generation rate at 14 days after shutdown is 6.38 MW1t1.

Basis: This input assumes that the core offload is complete and the plant is ready to initiate Decom Phase 2 at 14 days after reactor shutdown. The total spent fuel pool heat load at 14 days was determined in section 6.1.1 of this calculation. This assumption is conservative for all times after the reactor has been shut down for at least 14 days.

5.2.2. SFP Heat generation rate at 1 year after shutdown is 1.57 MW1t1.

Basis: This input assumes that all of the spent fuel remains in the spent fuel pool (no ISFSI campaigns). The total spent fuel pool heat load at 1 year was determined in section 6.1.1 of this calculation. This assumption is conservative for all times after the reactor has been shut down for at least 1 year.

5.2.3. Cooler flow rate (GPM) is 1000 GPM.

Basis: This is the design flow rate as described in UFSAR section 9.4.

5.2.4. "Design" NSCCW Inlet temperature (°F) is 95. The maximum "expected" NSCCW inlet temperature is 80°F.

Basis: The design NSCCW inlet temperature is described in UFSAR section 9.4 and 9.6.

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos.

202 I Sheet 13of66 The maximum expected NSCCW temperature is based on the planned shutdown no later than 30 September 2019. The river water temperature in October (or later) would be colder than the design temperature. The PPC data for RW inlet temperature (A0089) was reviewed for 2017 and 2016. River water temperatures varied between 56°F and 72°F in 2017, and between 54°F and 72°F in 2016. The maximum expected NSCCWtemperature of 80°F is based on (1) a conservative river water temperature for October of 75°F and (2) lower NSCCW to NR differential temperature (of 5°F) because the spent fuel pool will be the only significant heat load on the NSCCW/NR systems.

5.2.5. Spent Fuel Cooler HX Effectiveness is 0.516 at 140°F or above, and 0.514 between 130

& 140°F.

Basis: The design maximum spent fuel pool temperatures described in UFSAR section 9.4 were determined in Reference 3.11. The same analytical model used in Reference 3.11 (section 5) is used in Reference 3.10. Reference 3.10 Table 5.1 provides the "cooler effectiveness" coefficients used for the design analysis. Using a value of 0.516 is conservative for all temperatures at or above 140°F.

5.3. Purpose #3 Design Inputs 5.3.1. SFP Heat Generation rate calculated from section 6.1.1. All the fuel assemblies currently in the SFP and the core are assumed to stay in the pool indefinitely following permanent shutdown.

Basis: This is conservative because ISFSI campaigns will be scheduled and completed in the years following permanent shutdown to move the fuel assemblies from the spent fuel pool to dry cask storage. Each ISFSI campaign would remove additional heat loads from the pool resulting in higher TTB and TTAF times.

5.3.2. Mass of SFP water available for TTBITTAF is 3.43E 6 lbm (Reference 3.12).

Basis: The volume of SFP water available for TTB and TTAF is the water above the elevation of the fuel (318.75 ft) up to the low-level limit (343.5 ft). This is conservative because a large quantity of water is available for heat-up in the fuel region as well as the north end of the "A" pool being empty due to inaccessibility. The available mass for TTAF is different than Reference 3.12 available mass for boiloff because this analysis is calculating the time to top of active fuel (318.75 ft) while Reference 3.12 was calculating time to 7 feet above the fuel (325. 75 ft).

5.4. Purpose #4 Design Inputs See Attachment 1 Section 5.

5.5. Purpose #5 Design Inputs 5.5.1. The minimum uranium loading of the limiting fuel assembly in the Cycle 22 Core is 485. 75 kgU (Reference 3.4) .

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos.

202 I Sheet 14 of66 S.S.2. The atomic weight of uranium is 238.029 g/mole (Reference 3.16).

S.S.3. The atomic weight of oxygen is 1S.999 g/mole (Reference 3.16).

S.S.4. The density of MS is 6.SO g/cm 3 at 70°F (Reference 3.17).

S.5.S. Fuel Mechanical Data (Reference 3.1 S):

Guide Tubes

Number per bundle = 16

Material = MS

Guide Tube Length = 1S6.249 inches

Guide Tube ID= 0.498 inches

Guide Tube OD = O.S30 inches Instrument Tube

Number per bundle = 1

Material = MS

Instrument Tube Length (Reference 3.21) = 155.28S inches

Instrument Tube ID*= 0.400 inches

Instrument Tube OD= 0.493 inches Fuel Rods

Number per bundle = 208 rods

Material = MS

Fuel Rod Length (Reference 3.20) = 1SS.00 inches

Cladding OD = 0.430 inches

Cladding ID= 0.380 inches HTP Spacer Grids

Number per bundle = 7

Material = MS

Mass= 1088.7 grams HMP Spacer Grid

Number per bundle = 1

Material= lnconel Alloy 718

Mass= 1197.7 grams Upper End Fitting

Material = CF3 stainless steel

Mass**= 21.9 lbs Lower End Fitting

Material CF3 stainless steel

Mass**= 14.9 lbs

The MONOBLOC' Instrument Tube has a smaller inner diameter (0.352 inches) at the upper end of the tube (Reference 3.1S). Assuming the larger inner diameter is uniform over the active length of the fuel is conservative because it results in a smaller mass of cladding available for heat-up.

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos. , Sheet 202 15 of66

- The masses for the upper and lower end fittings include all small components such as stop pins, number plates, shims, spacers, cruciform springs, bolts, and nuts.

5.5.6. The initial fuel temperature is 110°F (316 K).

Basis: This value is conservative based on the maximum SFP temperature of 105.4°F determined in section 6.2.1 of this calculation. Setting the initial temperature to the maximum pool temperature at 1 year after shutdown is conservative because the spent fuel pool temperature at the minimum decay time of 488 days would be lower than at 1 year after shutdown, thus increasing the heat-up time.

5.5.7. The heat capacity for all fuel component materials are determined at the midpoint between the initial and final fuel temperatures, 881°F (745 K).

Basis: As shown in Attachment 2, the heat capacities for MS and U02 increase with respect to temperature. This means that, as the fuel assembly temperature increases, more time would be required to heat the fuel assembly mass to 900°C. Therefore, using the midpoint temperature for material properties is conservative with respect to the assembly heat-up.

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 I Rev. No.

1 I System Nos.

202 I Sheet 16 of66

6. Method of Analysis & Numerical Analysis 6.1. Purpose #1 Methodology 6.1.1. Post-Defueled SFP Heat Load The methodology described in ANSl/ANS-5.1 Decay Heat Power in Light Water Reactors was used to determine the total spent fuel pool heat load. The calculations were performed in a DTQSA controlled spreadsheet. A separate spreadsheet (Appendix 7.1) was created to yield the SFP heat load at various times after shutdown. This was then converted into a plot of decay heat vs. time after shutdown (Figure 2.1 ).

6.2. Purpose #2 Methodology 6.2.1 . Maximum SFP Temperature The methodology described in Reference 3.11 section 5.5 for determination of "Bulk Pool Temperature" is applied for this calculation. For the case of the maximum pool temperature the heat generated in the spent fuel equals the quantity of heat being transferred to the NSCCW system. In this case Equation 6.2.1 Where:

=

OsF heat generated in spent fuel, BTU/HR OsFc = heat transferred through spent fuel cooler, BTU/HR rh = flow rate through spent fuel cooler, lbm/hr

=

Hx heat exchanger effectiveness

=

T POOL bulk temperature in SF pool, °F T Ns = NSCCW temperature into Spent Fuel Cooler A spreadsheet (Appendix 7.2) was used to determine the maximum SF pool temperatures for one SF cooling train in service, both SF cooling trains in service, and "expected" NSCCW conditions at both 14 days and 1 year after shutdown.

6.3. Purpose #3 Methodology 6.3.1. Time to Boil The methodology described in Reference 3.14 Enclosure 6.13 for Time to Boil was used in this calculation. A spreadsheet (Appendix 7.3) was used to calculate the TIS at various times after shutdown and converted to a plot of TIS vs. time after shutdown (Figure 2.3.1 ).

Equation 6.3.1 Llt (TTB) = 2.93E- 7 * ( mass)

DH * (212 - Initial T)

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DECOM Spent Fuel Pool TH Analysis I Design Analysis IRev. No. I System Nos. I C-1101-202-E410-476 1 202 Sheet 17 of 66 Where:

at (TIB) = time to boil, hr

=

Mass Mass available for TIS, lbm DH = Decay Heat, MW1t1

=

Initial T Initial Temperature, °F 6.3.2. Time to Top of Active Fuel The methodology described in Reference 3.14 Enclosure 6.13 for Time to Boil off was used in this calculation. The calculation assumes that the spent fuel pool has already reached saturation. This time is conservative because it would take additional time for the pool to reach saturation before beginning to boiloff. A spreadsheet (Appendix 7.3) was used to calculate the TIAF at various times after shutdown and converted to a plot of TIAF vs.

time after shutdown (Figure 2.3.2).

Equation 6.3.2 L\t (TTAF) = ( 3.37E-4 * (~a~s)) + ( 24 ::y)

Where:

=

8t (TIAF) time to top of active fuel, days

=

Mass Mass available for TIAF, lbm

=

DH Decay Heat, MW1t1 6.4. Purpose #4 Methodology See Attachment 1 Sections 6 and 7.

6.5. Purpose #5 Methodology Section 6.5.10 determines the decay heat (q) required to raise the fuel assembly mass from the initial temperature to 900°C in 1O hours. This calculation conservatively ignores any heat transfer from the fuel assembly to the environment around the assembly and assumes that the heat-up time begins when the SFP is completely drained ignoring the time required to drain-down/boil-off the SFP. Equations in sections 6.5.2 and 6.5.7 below are used to develop a conservative mass for the U02 fuel, fuel rods, guide tubes, and instrument tube. The FA heat up rate is determined using Equation 6.5.10. The specific heat of Uranium Dioxide, MS, CF3 stainless steel, and lnconel Alloy 718 are determined in sections 6.5.3, 6.5.4, 6.5.5, and 6.5.6, respectively.

This decay heat generation rate determined in section 6.5.10 is used as an input into the equation determined in Attachment 1 section 7.2 to find the minimum decay time required to achieve this decay heat.

The methodology is recognized to be very conservative. The analysis does not credit the mass of the racks, and the spent fuel is loaded in the pool in a 1-in-5 pattern (

Reference:

RIN 3150-AJ59 "Regulatory Improvements for Power Reactors Transitioning to Decommissioning" Appendix A) .

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DECOM Spent Fuel Pool TH Analysis I Design Analysis 'Rev. No. , System Nos. , Sheet C-1101-202-E410-476 1 202 18of66 6.5.1. Temperature Rise The temperature rise for this analysis begins at the assumed initial temperature of 110°F (316 K) (Input 5.5.6) and ends at final temperature of 900°C (1173 K). Therefore, the temperature rise is:

Equation 6.5.1 LlT = Tfinal - Tinitial Using the values above yields the temperature rise:

LlT = 857 K 6.5.2. Mass and Atomic Weight of Uranium Dioxide Since the mass of the uranium in the uranium dioxide is known (Input 5.5.1), the total mass of uranium dioxide can be determined based on the chemical composition using Equation 6.5.2-1. The limiting fuel batch (22A) has 16 pins that contain gadolinium; however, the gadolinium is not included in this analysis. Neglecting the mass of the gadolinium is conservative because the additional thermal mass would increase the decay heat required to heat-up the assembly; thus, reducing the decay time.

Equation 6.5.2-1 U0 2 -+ U + 20 muo, = (mu + 2

mu~:)) *(1000 ;g)

Where:

muo2 = mass of uranium dioxide, g mu = mass of uranium, kg Ao= atomic weight of oxygen, g/mole Au = atomic weight of uranium, g/mole Inputting the above values yields the mass of uranium dioxide:

mu 02 = 5.51E 5 g The atomic weight of uranium dioxide is equal to the sum of the individual atomic weights of the uranium and oxygen.

Equation 6 .5. 2-2 Auo 2 =Au + 2

A 0 1!!j1;1eti1R EUi.2 2

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No. , System Nos. , Sheet 1 202 19of66 Where:

Auo2 = atomic weight of uranium dioxide, g/mole Ao = atomic weight of oxygen, g/mole Au = atomic weight of uranium, g/mole Inputting the above values yields the uranium dioxide atomic weight:

Auo 2 = 270.027 g/mole 6 .S.3. Specific Heat of Uranium Dioxide Uranium Dioxide specific heat is governed by the following equation (Reference 3.17):

Equation 6.S.3 for 298K s; Tk s; 3120K Where:

Cp, uo2 = specific heat of U02, J/g-K e =53S.8S K Eo = 1S7.7707*103 J/mol K1 = 80.1314 J/mol-K K2 = 32.845*10-4 J/mol-K2 K3 = 23.62183*10 6 J/mol R = 8.134 J/mol-K (ideal gas constant)

Tk =Temperature, K Auo2 = atomic weight of uranium dioxide, g/mole Inputting the above values yields uranium dioxide specific heat:

I Cp,U02 = 0.302 g - K 6.S.4. Specific Heat of MS Specific heat for MS is governed by the following equation (Reference 3.17):

Equation 6.S.4 Cp,MS = 0.2375 + 0.0001591

T for 273K s; T s; 1100K Where:

Cp, Ms= specific heat of MS, J/g-K

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476

'Rev. No. , System Nos. , Sheet 1 202 20of66 T =Temperature, K Inputting the above values yields the M5 specific heat:

J Cp MS = 0.356 --K

' g-6.5.5. Specific Heat of CF3 Stainless Steel CF3 stainless steel is thermodynamically the same as 304L stainless steel; therefore, the specific heat is governed by the same equation (Reference 3.18):

Equation 6.5.5 cp,ss = (0.1122 + 3.222x10-s

T) * (

4.184 cal J) 1 Where Cp,ss =specific heat of CF3 stainless steel, J/g-K T =Temperature, K Inputting the above values yields the CF3 stainless steel specific heat:

J Cp,SS = 0.570 g - K This value will be applied to the total mass of the upper and lower end fittings even though the masses include all the small components that may not be made up of CF3 stainless steel. The effects of these small components (shims, spacers, cruciform springs, bolts, and nuts) are considered inconsequential to the overall calculation due to their minimal masses.

6.5.6. Specific Heat of lnconel Alloy 718 The specific heat for lnconel Alloy 718 was found using the table in Appendix 7.5.

Interpolating and converting units (1 BTU/lb-F = 4.186798 J/g-K) yields the lnconel Alloy 718 specific heat, Cp,HMP:

J Cp,HMP = 0.583 g - K 6.5.7. M5 Volume The volume of M5 is determined by summing the volume of each of the credited M5 components (fuel rods, guide tubes, instrument tube) within the fuel assembly using the following equation. The M5 HTP grids are credited separately because their mass is known.

Equation 6.5.7

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No. , System Nos. , Sheet 1 202 21 of66 VMS.rods = (VFR + VGT + V1T) * ( 16.3871 cm3)in3 Where VMs,rods = total volume of MS rods, guide tubes, and instrument tube, cm 3 VFR = volume of MS in the fuel rods, in 3 VGr =volume of MS in the guide tubes, in 3 V1r = volume of MS in the instrument tube, in 3 Inputting the values calculated below yields a total MS volume:

VMS.rods = 18029.6 cm 3 6.5.7.1. Fuel Rods Equation 6.5.7.1 D~R,o - D~R.i)

VFR = ( TC

4

NFR

LFR Where DFR,o =fuel rod outer diameter, inches DFR,i = fuel rod inner diameter, inches NFR = number of fuel rods LFR = length of fuel rod, inches Inputting the above values yields a total fuel rod MS volume:

VFR = 1025.5 in 3 6.5.7.2. Guide Tubes Equation 6.5.7.2 Where DGT,o =guide rod outer diameter, inches DGr,i = guide tube rod inner diameter, inches NGT = number of guide tubes LGT = length of guide tube, inches Inputting the above values yields a total guide tube MS volume:

VGT = 64.6 in 3

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos.

202 I Sheet 22 of 66 6.5.7.3. Instrument Tube Equation 6.5.7.3 D~T.o - D~T.i)

V1T = ( 7t

4

NIT

L1T Where D1r.o =instrument tube outer diameter, inches D1r.i = instrument inner diameter, inches N1r = number of instrument tubes L1r = length of instrument tube, inches Inputting the above values yields an instrument tube M5 volume:

V1T = 10.1 in 3 6.5.8. Mass of HTP Spacer Grids The total mass of the MS spacer grids can be found by multiplying the known mass of an individual grid by the number of grids per bundle.

Equation 6.5.8 mMs,HTP = NHTP

ffittTPgrid Where mMs,HTP = total mass of HTP spacer grids, grams NHrP = number of HTP spacer grids per bundle MHrP grid = mass of a single HTP spacer grid, grams Inputting the above values yields a total HTP spacer grid mass:

ffiMS,HTP = 7620.9 g 6.5.9. Mass of End Fittings The total mass of the upper and lower end fittings can be found by summing the individual masses.

Equation 6.5.9 453.592 gram) mss = (muEF + mLEF) * ( l lb Where mss = total mass of the end fittings, grams muEF = mass of the upper end fitting, lbs mLEF = mass of the lower end fitting, lbs

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos.

202 I Sheet 23 of66 Inputting the above values yields the total mass of the end fittings:

mss = 16692.2 g 6.S.10. Decay Heat Generation Rate The required decay heat is determined using the thermal capacity of materials.

Equation 6.S.10-1 q= ( m

Cp

LltT) * ( 0.000278 J;J Where:

q = heat generation, W m = p*V =mass of material, g Cp = specific heat, J/g-K AT = temperature rise, K t = heat-up time, hr p =density, g/cm 3 V = volume, cm 3 For this analysis, there are four materials that are considered: uranium dioxide, MS, CF3 stainless steel, and lnconel Alloy 718. The fuels pellets are uranium dioxide. The fuel rod cladding, guide tubes, instrument tube, and HTP spacer grids are MS. The upper and lower end fittings are CF3 stainless steel. The lower HMP grid is lnconel Alloy 718. Under adiabatic conditions, the materials are modeled as heating up at the same rate consistent with Assumption 3 in Reference 3.22; therefore, AT/twill be the same for all materials.

This is justified because there is direct contact, thus conductive heat transfer, between the fuel rods and the lower end fitting and additional conductive heat transfer through the welded cage structure to both end fittings. Separating out the materials and substituting density and volume for MS rods results in the following equation:

Equation 6.S.10-2 AT q= (t * (muo 2

Cp,uo 2 + (PMs

VMs,rods + mMs,HTP)

Cp,Ms + mss

Cp,ss + ffittMP

Cp,HMP)) * ( 0.000278 J;J Where:

Xuo2 signifies the property is for uranium dioxide XMs signifies the property is for MS Xss signifies the property is for CF3 stainless steel XHMP signifies the property is for lnconel Alloy 718 Inputting the values above yields a required decay heat:

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No. , System Nos. , Sheet 1 202 24 of66 q = 5270W 6.5.11. Minimum Decay Time The decay heat determined using Equation 6.5.10-2 was input into the equation determined in Attachment 1 shown below.

Y (days)= 1.01 * [-9.22479E-09*x 3 + 1.59163E-04*x2 - 1.01332E+OO*x + 2.75333E+03]

Where:

Y = Decay Time, days x = Decay Heat, W Inputting the decay heat from section 6.5.10 yields the required decay time:

Y =488 days

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No. , System Nos. , Sheet 1 202 25 of66

7. Appendices 7.1. Purpose #1 Excel Calculations and Results 7.2. Purpose #2 Excel Calculations and Results 7.3. Purpose #3 Excel Calculations and Results 7.4. Purpose #5 Excel Calculations and Results 7.5. Specific Heats of lnconel 706 and lnconel 718 Alloys

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 I System Nos.

202 I 26Sheet of 66 Appendix 7.1 Purpose #1 Excel Calculations and Results Time after SFP Decay Time after SFP Decay Time after SFP Decay Shutdown Heat Shutdown Heat Shutdown Heat

{days} {MWth} {days} {MWth} {days} {MWth}

14 6.38 504 1.32 994 0.87 28 4.89 518 1.30 1008 0.87 42 4.16 532 1.28 1022 0.86 56 3.70 546 1.26 1036 0.85 70 3.38 560 1.25 1050 0.84 84 3.14 574 1.23 1064 0.84 98 2.94 588 1.21 1078 0.83 112 2.78 602 1.20 1092 0.82 126 2.64 616 1.18 1106 0.82 140 2.51 630 1.16 1120 0.81 154 2.41 644 1.15 1134 0.80 168 2.31 658 1.13 1148 0.80 182 2.22 672 1.12 1162 0.79 196 2.14 686 1.11 1176 0.78 210 2.07 700 1.09 1190 0.78 224 2.00 714 1.08 1204 0.77 238 1.94 728 1.07 1218 0.77 252 1.89 742 1.05 1232 0.76 266 1.84 756 1.04 1246 0.75 280 1.79 770 1.03 1260 0.75 294 1.75 784 1.02 1274 0.74 308 1.70 798 1.01 1288 0.74 322 1.67 812 1.00 1302 0.73 336 1.63 826 0.98 1316 0.73 350 1.60 840 0.97 1330 0.72 364 1.57 854 0.96 1344 0.72 378 1.54 868 0.95 1358 0.71 392 1.51 882 0.94 1372 0.71 406 1.48 896 0.94 1386 0.71 420 1.46 910 0.93 1400 0.70 434 1.43 924 0.92 1414 0.70 448 1.41 938 0.91 1428 0.69 462 1.39 952 0.90 1442 0.69 476 1.37 966 0.89 1456 0.69 490 1.34 980 0.88

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No. , System Nos. , Sheet 1 202 27of66 Appendix 7 .2 Purpose #2 Excel Calculations and Results Design Design Maximum Maximum (both Maximum (only "expected" pool SFP trains) one SFP train) temperature SFP Heat generation rate (MW) 6.38 6.38 6.38 SFP Heat generation rate (BTU/HR) 21768560 21768560 21768560

of cooler trains in service 2 1 2 Cooler flow rate (GPM) 1000 1000 1000 Cooler flow rate (LBM/HR) 499800 499800 499800 NSCCW Inlet temperature (F) 95 95 80 HX Effectiveness 0.514 0.514 0.514 Pool water temperature (F) 137.4 179.7 122.4 Design Design Maximum Maximum (both Maximum (only "expected" pool SFP trains) one SFP train) temperature SFP Heat generation rate MW 1.57 1.57 1.57 SFP Heat generation rate BTU/HR 5356840 5356840 5356840

of cooler trains in service 2 1 2 Cooler flow rate (GPM) 1000 1000 1000 Cooler flow rate (LBM/HR) 499800 499800 499800 NSCCW Inlet temperature (F) 95 95 80 HX Effectiveness 0.514 0.514 0.514 Pool water temperature (F) 105.4 115.9 90.4

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 I Rev. No. , System Nos. , Sheet 1 202 28 of66 Appendix 7 .3 Purpose #3 Excel Calculations and Results Time after SFP Decay Time after SFP Decay Shutdown Heat TTB TTAF Shutdown Heat TTB TTAF

{da~s) {MWth} {hrs} {da~s} (days} {MWth} {hrs} {days}

14 6.38 13.71 7.55 504 1.32 66.10 36.41 28 4.89 17.87 9.84 518 1.30 67 .13 36.98 42 4.16 21.04 11 .59 532 1.28 68.14 37.54 56 3.70 23.65 13.03 546 1.26 69.15 38.09 70 3.38 25.88 14.26 560 1.25 70.16 38.65 84 3.14 27.87 15.35 574 1.23 71 .16 39.20 98 2.94 29.71 16.37 588 1.21 72.15 39.74 112 2.78 31.47 17.33 602 1.20 73.14 40.29 126 2.64 33.15 18.26 616 1.18 74.12 40.83 140 2.51 34.77 19.16 630 1.16 75.10 41.37 154 2.41 36.35 20.03 644 1.15 76.08 41.91 168 2.31 37.89 20.87 658 1.13 77.05 42.44 182 2.22 39.39 21.70 672 1.12 78.01 42.97 196 2.14 40.85 22.50 686 1.11 78.98 43.50 210 2.07 42.27 23.28 700 1.09 79.95 44.04 224 2.00 43.64 24.04 714 1.08 80.95 44.59 238 1.94 45.00 24.79 728 1.07 81.95 45.14 252 1.89 46.33 25.52 742 1.05 82.94 45.69 266 1.84 47.62 26.23 756 1.04 83.92 46.23 280 1.79 48.88 26.92 770 1.03 84.90 46.77 294 1.75 50.10 27.60 784 1.02 85.88 47.31 308 1.70 51.30 28.26 798 1.01 86 .85 47 .84 322 1.67 52.46 28.90 812 1.00 87.82 48 .37 336 1.63 53.60 29.52 826 0.98 88.78 48.90 350 1.60 54.71 30.14 840 0.97 89.73 49.43 364 1.57 55.80 30.74 854 0.96 90.68 49.95 378 1.54 56.87 31.33 868 0.95 91.63 50.47 392 1.51 57.92 31.91 882 0.94 92.56 50.99 406 1.48 58.96 32.48 896 0.94 93.49 51.50 420 1.46 59.98 33.04 910 0.93 94.41 52.01 434 1.43 61.00 33.60 924 0.92 95.32 52.51 448 1.41 62.00 34.15 938 0.91 96.27 53.03 462 1.39 63.00 34.70 952 0.90 97.22 53.56 476 1.37 64.04 35.27 966 0.89 98.17 54.07 490 1.34 65.07 35.85 980 0.88 99.10 54.59

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos.

202 I Sheet 29of66 Time after SFP Decay Shutdown Heat TTB TTAF (da~s} (MWth) (hrs) (days) 994 0.87 100.03 55.10 1008 0.87 100.96 55.61 1022 0.86 101.87 56.12 1036 0.85 102.78 56.62 1050 0.84 103.68 57.11 1064 0.84 104.58 57.61 1078 0.83 105.46 58.09 1092 0.82 106.34 58.58 1106 0.82 107.20 59.05 1120 0.81 108.06 59.53 1134 0.80 108.91 60.00 1148 0.80 109.76 60.46 1162 0.79 110.61 60.93 1176 0.78 111.51 61.42 1190 0.78 112.39 61.91 1204 0.77 113.27 62.40 1218 0.77 114.14 62.87 1232 0.76 115.00 63.35 1246 0.75 115.85 63.82 1260 0.75 116.69 64.28 1274 0.74 117.53 64.74 1288 0.74 118.35 65.19 1302 0.73 119.17 65.64 1316 0.73 119.97 66.09 1330 0.72 120.76 66.52 1344 0.72 121.54 66.95 1358 0.71 122.32 67.38 1372 0.71 123.08 67.80 1386 0.71 123.84 68.21 1400 0.70 124.58 68.63 1414 0.70 125.32 69.03 1428 0.69 126.05 69.43 1442 0.69 126.77 69.83 1456 0.69 127.48 70.22

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DECOM Spent Fuel Pool TH Analysis I Design Analysis IRev. No.

C-1101-202-E410-476 1 ISystem Nos. I Sheet 202 30 of 66 Appendix 7.4 Purpose #5 Excel Calculations and Results Initial Temperature 110 F 43 c 316 K Final Temperature 1652 F 900 c 1173 K Delta Temperature 857 K Specific Heat Temperature 881 F 472 c 745 K Mass of U02 Mass of U 485.75 kgU Atomic Mass of U 238.029 g/gmole Atomic Mass of 0 15.999 g/gmole Mass of O 65.30 kg02 Mass of U02 551.05 kaU02 Mass of U02 5.51E+05 gU02 Molar Mass U02 270.027 a/a mole Heat Capacity (U02) e 535.85 K E_D 157770.7 J/mol K_1 80.1314 J/mol-K K_2 3.2845E-03 J/mol-K K_3 2.362183E+07 J/mol R 8.314 J/mol-K Temperature 745 K Molar Mass 270.027 g/mol Specific Heat 81.7 J/mol-K I Heat Capacity 0.302 J/g-K I

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No. , System Nos. , Sheet 1 202 31 of 66 Heat Capacity of M5 l Heat Capacity 0.356 J/g-K Heat Capacity of Alloy 718 l Heat Capacity 0.5830 J/g-K Heat Capacity of CF3 stainless steel l Heat Capacity 0.570 J/g-K Volume of M5 Fuel Rods Value Units Number of Rods 208 Inner Diameter 0.380 in Outer Diameter 0.430 in Fuel Rod Length 155.00 in I Volume 1025.5 in 3 Guide Tubes Value Units Number of Tubes 16 Inner Diameter 0.498 in Outer Diameter 0.530 in GT Assembly Length 156.249 in I Volume 64.6 in 3 Instrument Tube Value Units Number of Tubes 1 Inner Diameter 0.400 in Outer Diameter 0.493 in Length 155.285 in I Volume 10.1 in 3 VolumeofM5 1100.2 in 3 Volume of M5 18029.6 cm 3

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

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202 I Sheet 32 of66 MS Spacer Grids Number of Grids 7 Mass of each grid 1088.7 ~

I Total Mass 7620.9 HMP Spacer Grid Number of Grids 1 Mass of Each Grid 1197.7 ~

I Total Mass 1197.7 Upper End Fitting Mass 21.9 lbs I Mass 9933.7 g Lower End Fitting Mass 14.9 lbs I Mass 67S8.S2 g Total Mass of End Fittings Mass of UEF 9933.7 g Mass of LEF 67S8.S ~

I Total Mass Decay Heat Delta Temperature 8S6.7 K Time 10 hrs Mass of U02 S.S1E+OS gU02 Heat Capacity of U02 0.302 J/g-K Density of MS 6.SO g/cm 3 Volume of MS 18029.S71 cm 3 Mass of MS Spacer Grids 7620 .9 g Heat Capacity of MS 0.3S6 J/g-K Mass of End Fittings 16692.186 g Heat Capacity of CF3 O.S70 J/g-K Mass of HMP Grid 1197.7 g Heat Capacity of Alloy 718 O.S83 J/g-K I Decay Heat S270 w

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DECOM Spent Fuel Pool TH Analysis I Design Analysis 'Rev. No. , System Nos. , Sheet C-1101-202-E410-476 1 202 33 of66 Decay Time Constants -9.22479E-09 x3 1.59163E-04 x2

-1.01332E+OO x

2. 75333E+03 I Decay Time 488 days

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 I Rev. No.

1 ISystem Nos. , Sheet 202 34 of66 Appendix 7.5 Specific Heats of lnconel 706 and lnconel 718 Alloys TABLE 5 SPECIFIC HEATS OF INCONEL 706 AND INCONEL 718 ALLOYS Temperature (OF) Specific Heat (Btu lb-1 F-1)

Inconel 706 Inconel 718 70 0.111 0.107 100 0.112 0.109 150 0.114 0.112 200 0.116 0.114 2fi0 0.117 0.116 300 0.119 0.119 350 0.120 0.121 400 0.121 0.123 450 0.122 0.125 500 0.123 0.126 550 0.123 0.128 600 0.124 0.130 650 0.125 0.131 700 0.125 0.133 750 I) .126 0.135 800 0.128 0.137 850 0.132 0.138 900 0.136 0.140 950 0.141 0.142 1000 0.144 0.143 1050 0.148 0.145 1100 0.149 0.149 1150 0.150 0.160 1200 0.151 0.166 1250 0.151 0.169 1300 0.150 0.169 1350 0.149 0.169 1400 0.148 0.168 1450 0.147 0.167 1500 0.146 0.166

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 I Rev. No. , System Nos. ,

1 202 Sheet 35 of66 ATTACHMENT 1 ORIGEN2 Decay Heat Calculation for Fuel Assembly with Maximum Decay Heat Load 1.0 PURPOSE To calculate the decay heat from the fuel assembly determined to produce the highest decay heat load of all fuel assemblies in the TMI Spent Fuel Pool (SFP) between 1 and 3 years after shut down of TMl-1 Cycle 22. The resultant decay heat load is to be used as input to the Beyond Design Basis zirconium fire event for a drained SFP. Best estimate decay heat loads will be calculated using the ORIGEN2 code based on reactor rated power of 2568 MWt and a Cycle 22 length of 720 EFPD, which is the maximum licensed length for Cycle 22.

2.0

SUMMARY

OF RESULTS A best estimate decay heat load for the fuel assembly with the maximum heat load in the TMI SFP following TMl-1Cycle22 shutdown has been calculated using the ORIGEN2 computer code.

Per Section 7.3 of this calculation, the decay heat load from this Batch 22A fuel assembly as a function of decay time (in days) after Cycle 22 shutdown is shown in the table below. This decay heat load is appropriate to be used as input to the Beyond Design Basis zirconium fire event for a drained SFP.

Maximum Fuel Assembly Decay Heat (Watts} for Various Decay Times After TMl-1 Cycle 22 Shutdown Decay Time (Days) 365.0D 548.0D 730.0D 913.0D 1095.0D Decay Heat (W) 6.44E+o3 4.70E+o3 3.65E+o3 2.94E+o3 2.44E+o3 The following polynomial fit represents this data for decay heat (x) between 2.44 and 6.44 kW.

The polynomial fit is in good agreement with the individual ORIGEN2 data points. A 1% multiplier is applied to the Excel fit for conservatism.

Y (days}= 1.01*[-9.22479E-09*x3 +1.59163E-04*x2 -1.01332E+OO*x + 2.75333E+03]

3.0 REFERENCES

3.1 RSIC Code Package CCC-371, "ORIGEN 2.1, Isotope Generation and Depletion Code Matrix Exponential Method," May 1999.

3.2 ORNUTM-11018, "Standard- and Extended-Burnup PWR and BWR Reactor Models for the ORIGEN2 Computer Code," S. Ludwig, J. Renier, December 1989.

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

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202 I Sheet 36of66 3.3 AREVA Calculation FS1-0030101, Rev. 1, "TMl-1 Cycle 22 FFCD," 2/03/2017.

3.4 Exelon Nuclear Fuels TMI Fuel Database, "C22database.accdb," 1/03/2018.

3.5 Exelon TOOi NF173266, Rev. 0, "TMI End-of-Cycle 21 Data and FIDMS Files," 9/18/2017.

3.6 Exelon DTSQA document, "ORIGEN Version 2.1, EX0004724 Release Notes," October 2014.

3.7 AREVA Doc. FS1-0029697, Rev. 1, "Task 76 Inputs, Mechanical Design Information for Safety Analysis, and Regulatory Compliance for TMl1-22," 12/19/2016.

3.8 AREVA Doc. 8AW-10227P-A, Rev. 1, "Evaluation of Advanced Cladding and Structural Material (MS) in PWR Reactor Fuel," June 2003.

3.9 ASTM Specification 8637 for Alloy 718 (see Appendix 8) 4.0 ASSUMPTIONS 4.1 Fuel batch burnups are taken from NEMO calculations documented in Tables 38-39 and Figure 34 of Reference 3.3. NEMO is the design code for TMl-1 core simulations and calculated burnups are representative of actual burn ups measured by the core monitoring system. 8urnups prior to Cycle 21 are based on core follow calculations that are depleted to actual cycle lengths. Cycle 21 burnups are based on a design depletion to 647 EFPD; differences from the actual Cycle 21 length of 648.8 EFPD (Reference 3.5) are negligible with respect to this decay heat calculation. Cycle 22 burnups are based on 720 EFPD, which is the maximum licensed cycle length for Cycle 22. Higher fuel assembly burnups are conservative for decay heat calculations.

4.2 Reference 3.3 lists fuel batch enrichments based on the nominal base enrichment of the batch. Since lower enrichments yield higher decay heat results, actual enrichments based on NRC 741 forms are used in this calculation. This includes segments of the fuel assemblies that have lower than nominal enrichments (i.e., Gad rods and axial blankets).

The actual enrichments were obtained from Reference 3.4.

4.3 For fuel burned in more than one cycle, ORIGEN2 runs ignored refueling outages. This has no impact on short-lived isotopes which reach equilibrium concentrations shortly after cycle startup and has a conservative, albeit minimal, impact on long-lived isotopes which continually increase in concentration as a function of exposure; ignoring intermediate decay periods will increase the final concentrations.

4.4 The ORIGEN2 cross-section library, PWRUE.Ll8, is used in this calculation as this is most representative of TMl-1 two-year cycles. The library is based on an "extended cycle" reactor model where fuel achieves 50 GWd/mtU burnup in three cycles.

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

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202 I Sheet 37of66 4.5 The decay heat loads calculated herein using ORIGEN2 are to be used as input to a Beyond Design Basis analysis. Therefore, ORIGEN2 results can be used directly as best estimate values, without applying any additional uncertainty factor.

4.6 Only fuel operating in Cycle 22 needs to be evaluated for maximum decay heat loads following Cycle 22 shut down. All other fuel in the SFP will have decayed for~ 2 years with a significant reduction in decay heat.

4.7 Only structural materials in the active fuel region receive sufficient flux to generate activation products with any appreciable decay heat load with respect to this calculation .

5.0 DESIGN INPUT 5.1 Batch-average burnups for fuel loaded in Cycle 22 are taken from Tables 38-39 and Figure 34 (for Batch 24A only) of Reference 3.3. The tables also contain the effective full power days (EFPD) of operation for all TMI cycles.

5.2 Actual batch enrichments for fuel loaded in Cycle 22 are taken from Reference 3.4. For each batch, the lowest as-built enrichment of all fuel assemblies in the batch is used to represent the batch.

5.3 Uranium loading for each fuel batch in Cycle 22 is taken from Table 14 of Reference 3.3.

5.4 Fuel assembly structural materials and dimensional data are taken from Reference 3.7.

5.5 Material specifications for M5 and Alloy 718 are taken from References 3.8 and 3.9, respectively. The Alloy 718 specification is copied to Appendix B; the maximum Cobalt content is assumed and elements less than 0.50% are ignored.

6.0 OVERALL APPROACH AND METHODOLOGY A fuel assembly's decay heat is a function of the fuel's power level (which affects the equilibrium concentration of both long and short-lived isotopes) and the exposure of the fuel (which affects long-lived isotopes that continue to accumulate). Maximizing these parameters will result in a higher decay heat load. Fuel assembly U235 enrichment has a second order effect on decay heat, with lower enrichments resulting in higher decay heat loads.

The fuel batches in Cycle 22 will be examined to determine which batches can be eliminated based on comparison of the three parameters of interest. The remaining batches will be evaluated using ORIGEN2 to calculate decay heat loads for various decay times after Cycle 22 shutdown.

The batch with the highest decay heat loads after 1 year of decay (expected minimum time for

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

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202 I Sheet 38 of66 the zirc fire analysis) will be selected to provide the maximum fuel assembly decay heat loads to the downstream zirc fire analysis.

Since the purpose of this calculation is to identify the fuel assembly with the highest heat load after 1 year of decay, batches that reside in only one eighth-core symmetric location (typically batches containing only 1, 4, or 8 fuel assemblies) can be treated as essentially the same. Larger batches will be examined to determine if one of the eighth-core symmetric locations for that batch has a higher burnup (and therefore specific power) than the others.

6.1 Computer Codes The DOS-based ORIGEN2 code, Version 2.1 (DTSQA application EX0004724) is used to calculate decay heat. The code was run on the nfw-ksq-01 Windows 2008 virtual server where it was approved for release in Reference 3.6. DOSBox was used to run ORIGEN2 on the Windows server with D:\dosprogs mounted as the D: drive.

The dto.bat batch file used to execute ORIGEN2 for this calculation provides the paths and filenames of the executable program and libraries that were called. A typical example of this batch file is listed in Appendix A; only the batch ID was changed in the input/output filenames for the various batches analyzed.

All cases were run in directory C:\Users\u001 rpj\ORIGEN21\TMl\DTO.

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476

'Rev. No. , System Nos. , Sheet 1 202 39 of66 7.0 CALCULATIONS 7.1 Comparison of Fuel Batches Operating in TMl-1 Cycle 22 The fuel batches operating in TMl-1 Cycle 22 are listed in Table 14 of Reference 3.3, which is copied below.

Table 14: TMl*1 Cycle 22 Fuel Inventory Base Gadolinia Number Design Batch ID Pin Layout FA Design Enrichment lo/ol Pins ofFA's Loadina fkaU) 19E3 4.95 12 x2.0 Fioure 12 Mark-B-HTP 8 489.18 20A3 1.40 None NIA Mark-B-HTP-1 1 489.83 12 x3.0 22A1 4.57 Figure 13 Mark-B-HTP-1 4 485.75 8 x8.0 12x2.0 2282 4.76 Figure 14 Mark-B-HTP-1 8 486.07 8 x8.0 22D 4.90 12 x2.0 Fioure 15 Mark-B-HTP-1 8 487.77 22E2 4.90 16 x2.0 Figure 16 Mark-B-HTP-1 8 487.56 23A2 4.10 None Figure 17 Mark-B-HTP-1 4 488.43 12 x 3.0 238 4.10 Figure 18 Mark-B-HTP-1 12 485.75 8 x8.0 23C 4.30 None Figure 17 Mark-B-HTP-1 8 488.43 16 x2.0 23D 4.30 Figure 19 Mark-B-HTP-1 4 486.91 4 x6.0 23E 4.50 None Figure 17 Mark-B-HTP-1 8 488.43 23F 4. 50 8 x2.0 Figure 20 Mark-B-HTP-1 16 487.99 23G 4.50 16 x2.0 Figure 21 Mark-B-HTP-1 8 487.56 12 x2.0 23H 4.50 Figure 22 Mark-B-HTP-1 8 486.07 8 x8.0 12 x3 .0 4.36 Figure 23 Mark-B-HTP-1 16 485.75 24A 8 x8.0 248 4.75 None Figure 17 Mark-B-HTP-1 4 488.43 24C 4.75 8 x2.0 Figure 24 Mark-B-HTP-1 12 487.99 240 4.75 16 x2.0 Figure 25 Mark-B-HTP-1 16 487.56 8 x3.0 4.75 Figure 26 Mark-B-HTP-1 8 486.07 24E 8 x8 .0 24F 4.88 8 x2.0 Figure 27 Mark-B-HTP-1 8 487.99 24G 4.88 8 x3.0 Figure 28 Mark-B-HTP-1 8 487.78 The batch average burnups for these batches are listed in Tables 38 and 39 from Reference 3.3, and are copied below. Also, EOC-22 quarter-core assembly burnups from Figure 34 of Ref. 3.3 are copied below. Inspection of these tables and figure allows the following batches in Cycle 22 to be eliminated from consideration:

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 I Rev. No.

1 I System Nos. , Sheet 202 40 of66

Batch 20A3 - total burnup and final cycle power level considerably lower than other batches

Batch 22E2 - total burnup and enrichment bounded by Batch 22A1

- From Figure 34, Batch 22E2 consists of quarter-core locations, 0-13 & H-15.

Location 0-13 has a slightly higher burnup (51,783 MWd/mtU) than Batch 22A1 (51,255 MWd/mtU). However, the enrichment difference (0.33 w/o) is considered more significant than the small burnup difference, especially after 1 year of decay.

Batches 23B thru 23H - total burn ups less than or equivalent to Batch 23A2, and enrichment bounded by Batch 23A2

- From Figure 34, Batch 23B consists of two eighth-core locations, H-12 & N-14.

Burnups from both locations are bounded by Batch 23A2.

- Batch 23E consists of two quarter-core locations, H-1 O & H-11. Location H-1 O has a slightly higher burnup (49,681 MWd/mtU) than Batch 23A2 (48,504 MWd/mtU).

However, the enrichment difference (0.40 w/o) is considered more significant than the small burnup difference, especially after 1 year of decay.

- Batch 23F consists of two quarter-core and one eighth-core locations, M-11, N-12

& M-12. Burnups from all locations are bounded by Batch 23A2.

Batches 24B thru 24G - total burn ups less than or equivalent to Batch 24A, and enrichment bounded by Batch 24A

- From Figure 34, Batch 24A consists of two eighth-core locations, K-12 & L-11. The higher burnup from L-11 (28,667 MWd/mtU) will be used for Batch 24A.

- Batch 24C consists of one quarter-core and one eighth-core locations, H-09 & K-

10. Burnups from both locations are bounded by Batch 24A.

- Batch 240 consists of two eighth-core locations, K-14 & L-14. Burnups from both locations are bounded by Batch 24A.

- Batch 24F has a burnup equivalent to Batch 24A, but has a uranium loading -0.5%

higher. Since fuel assembly specific powers are determined by multiplying burnup by uranium loading, Bach 24F would have a specific power -0.5% higher than Batch 24A. However, the enrichment difference (0.52 w/o) is considered more significant than the small power difference, especially after 1 year of decay.

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos. , Sheet 202 41of66 Table 38: TMl-1 Fuel Burnup (MWd/mtU) Distribution by Batch ID (17-20 Only) and by Cycle Number Rated Thermal Power= 2568 MWth Base 547111 120111 EFPD Batch of FA 679.16 694.90 700.30171 628.31 685.24 690.35 Enrichment ID by Cycle Cycle Cycle Cycle Cycle Cycle Cycle Cycle Batch-(%)

Batch 15 16 17 18 19 20 21 22 ~veraae 17A1.., 4.45 28 26707 21323 48030 1781..1 4.78 8 23197 24155 6303 53655 7C1 1.. 1 4.78 8 26876 22787 49663 11c2 1*' 4.78 8 26056 9407 12701 48164 1101 1* 4.90 1 23179 23271 46450 17021.. 1 4.90 15 23405 19993 10925 54323 18Al.ll 4.45 4 26777 17802 12985 57564 188 1'"1 4.45 16 27260 22236 49496 18C1'"1 4.60 12 27142 21999 49141 1801 '.. 4.60 8 27804 21564 49368 1802'" 4.60 8 25837 9196 18546 53579 18E1 '"1 4.80 12 23275 22796 46071 18E21"1 4.80 4 23819 12028 18198 54045 18F*' 4.80 8 23045 24305 47350 19A'"' 4.60 16 27927 19538 47465 198 1'"1 4.70 4 27524 19099 46623 l9C1 '"1 4.70 8 27336 20456 47792 19C21" 1 4.70 8 26105 8921 12206 47232 1901..1 4.80 12 28056 14701 42757 19E1 1.11 4.95 4 24459 11262 35721 19E2'"' 4.95 12 23860 21022 8025 52907 19E31"' 4.95 8 23522 22570 6562 52654 20A1 1.40 18 6549 10075 16624 20A2 1.40 1 6323 6141 15749 28213 20A3 1.40 1 6323 6141 17578 30042 208 2.50 1 17426 19258 15683 52367 20C1 1'" 4.60 4 24462 23056 47518 2oc2 1.. 4.60 4 24404 22101 7349 53854 200 1 ~ 1 4.60 4 241 78 22811 46989 l20E1 1.11 4.60 8 24943 22933 47876 20E2'"1 4.60 8 24178 8805 12600 45583 2op** 4.70 8 24385 23583 47968 20Gl-11 4.70 8 24389 22684 47073 20H1 '" 4.95 4 22246 23538 45784 20H2 1" 4.95 4 22281 22019 8718 53018 2oa1.11 4.95 8 22476 23262 7851 53589 20J\.>I 4.95 8 21869 24576 6424 52869

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 I System Nos. , Sheet 202 42of66 Table 39: TMl-1 Fuel Burnup (MWd/mtU) Distribution by Batch ID (21-24 Only) and by Cycle Rated Thermal Power = 2568 MWth Base Number Batch Enrichment of FA 68524 690.35 64J<61 720l61 --- --- EFPD ID Batch-(%) by Batch Cycle 19 Cycle 20 Cycle 21 Cycle 22 Cycle 23 Cycle 24 Average 21A1->' 4. 10 4 26531 21650 48181 1 4.40 12 27083 21955 49038 218 "'

21C1 t"'1 4.40 4 26579 21179 47758 21C2 1"J 4.40 12 26943 21976 48919 21C3l3l 4.40 8 26528 21749 5434 53711 210 1,,, 4.82 8 25790 8980 11941 46711 21E1 l"I 4.90 4 23015 23614 46629 21E2l"l 4.90 4 23022 19643 7366 50031 21F1 1" 1 4.95 4 24089 22407 46496 21F21"' 4.95 12 22822 22815 7593 53230 22A1 1" 4.57 4 27343 23912 51255 Z2A2 1..i 4.57 12 27299 21254 48553 22B1 1"i 4.76 12 27334 21305 48639 l22B2 1"i 4.76 8 25806 8421 13021 47248 22Ct"1 4.76 12 27555 21510 49065 220 1" 1 4.90 8 24541 21524 7929 53994 22E1 1..i 4.90 8 23090 22355 45445 22E2l"I 4.90 8 21991 19839 8520 50350 23A1 4. 10 4 24244 24244 23A2 4.10 4 24241 24263 48504 23Bt"1 4. 10 12 25063 13031 38094 23C 4.30 8 24528 24138 48666 23Dl.ll 4.30 4 25789 18251 44040 23E 4.50 8 25571 23385 48956 23F 1"' 4.50 16 20958 23659 44617 23Gt"f 4.50 8 22388 24549 46937 23Hl.>' 4.50 8 24083 23750 47833 24Al.ll 4.36 16 28400 28400 248 4.75 4 2n11 27711 24Cl"J 4.75 12 28429 28429 240£3 ) 4.75 16 24279 24279 24E1s1 4.75 8 26986 26986 24Fl"f 4.88 8 28473 28473 24G 1Jt 4.88 8 22845 22845

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

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202 I Sheet 43 of66 Figure 34: TMl-1 Cycle 22 Assembly-Average Burnup Distribution (GWd/mtU) 8 9 10 11 12 13 14 15 20A3 24C 23E 23E 238 248 230 22E2 30.042 28.466 49.68 1 48.232 46.860 27.71 1 44.040 48.918 H

12.464 0.000 25.578 25.565 24.735 0.000 25.789 40.910 17.578 28.466 24.103 22.667 22.125 27.711 18.251 8.008 24C 22A1 24C 23H 24A 23G 240 220 28.466 51 .255 28.408 47.853 28.250 46.898 24.874 53.955 K

0.000 27.343 0.000 24.083 0.000 22 .308 0.000 46 .020 28.466 23.912 28.408 23.770 28.250 24 .590 24.874 7.935 23E 24C 23A2 24A 23C 24F 240 19E3 49.681 28.413 48.504 28.667 48.708 28.494 23.705 52.632 L

25.578 0.000 24.241 0.000 24.510 0.000 0.000 46.065

24. 103 28.413 24.263 28.667 24.198 28.494 23.705 6.567 23E 23H 24A 23F 23F 24E 2282 48.232 47 .811 28.516 45.859 44.537 26.967 47.28 1 M 25.565 24.082 0.000 21 .447 20.474 0.000 34.279 22.667 23.729 28.516 24.412 24.063 26.967 13.002 238 24A 23C 23F 23F 24G 238 46.860 28.166 48.624 44.480 43.591 22 .782 33.781 N

24.735 0.000 24.547 20 .502 21.408 0.000 25.335

22. 125 28.166 24.077 23.978 22.183 22.782 8.446 248 23G 24F 24E 24G 22E2

27. 711 46.977 28.452 27 .005 22.908 51.783 0

0.000 22.468 0.000 0.000 0.000 42.751 27.711 24.509 28.452 27 .005 22.908 9 .032 230 240 240 2282 238 Batch ID p 44.040 24.846 23.692 47.214 33.640 720.0 EFPD 25.789 0.000 0.000 34.174 25.1 21 O.OEFPD 18.251 24.846 23.692 13.040 8.519 Delta BU 22E2 220 19E3 48.918 54.035 52.677 R 40.910 46.111 46.120 8.008 7.924 6.557

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

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202 I Sheet 44 of66 7.2 Structural Materials for ORIGEN2 Input Per assumption 4.7, only structural materials in the active fuel region will be modelled in ORIGEN2. This consists of fuel cladding, guide tubes, instrument tube, and spacer grids.

Quantities, materials, and dimensions for these components are taken from Ref. 3.7, as are spacer grid weights. The length for fuel rods, guide tubes, and instrument tube are set equal to the active fuel length of 143 inches. The density for MS material is taken from Ref 3.8, page A-3. These parameters are shown in the following table, along with calculated volumes and weights.

Volume (cc) Weight(g)

Description Quantity Material Density (g/cc) Dimensions Total weight (g) per unit per unit HTP Grids 7 MS --- --- --- 1088.7 7620.9 HMP Grids 1 Alloy718 --- --- --- 1197.7 1197.7 ID=0.380" Fuel Rods 208 MS 6.48 OD=0.430" 74.S 483.0 100,466.3 Length= 143.0" ID=0.498" Guide Tubes 16 MS 6.48 OD=O.S30" 60.S 392.3 6,277.2 Length= 143.0" ID=0.400" Instrument Tube 1 MS 6.48 OD=0.493" 1S2.8 990.S 990.S Length= 143.0" In the table above, tube volumes are calculated using the following formula:

2 2 Vo l ume ( cc) = ( 7t*(OD*2.54) -

4 7t*(ID*2.54) 4

)

(L eng th

2.54)

Tube weights are calculated using the formula:

Unit Weight (g) =Volume

Density Total weights are calculated using the formula:

Total Weight (g) =Unit weight* Quantity

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DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 I System Nos. ,

202 Sheet 45 of 66 Based on the table above, the total weight of MS and Alloy 718 in the active fuel region is:

11S,3S4.8S g of MS 1197.7 g of Alloy 718 The following MS alloy nominal chemistry is taken from Ref. 3.8, page A-3. Minor impurities are ignored.

MS 98.875% Zr 1.00% Nb 0.125% 0 The following Alloy 718 nominal chemistry is taken from Ref. 3.9 (see Appendix 8). Minor impurities are ignored.

Alloy 718 52.50% Ni 19.00°/o Cr 5.125% Nb 3.05% Mo o.90% n 0.50% Al 1.00% Co 17.925% Fe Multiplying the total weights of MS and Alloy 718 by their constituent elements, the following elemental weights are calculated.

M5(g) Alloy718(g) 114057.1 Zr 628.8 Ni 1153.5 Nb 227.6 Cr 144.2 0 61.4 Nb 36.5 Mo 10.8 n 6.0 Al 12.0 Co 214.7 Fe Combining elemental weights yields the following totals for ORIGEN2 input:

/

Exelonw DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No. , System Nos.

1 202 I Sheet 46 of66 Totals for ORIGEN2 Weight (g) Atomic#

144.19 0 8 Ad~ to weight of Oxygen from U02 5.99 Al 13 10.78 Ti 22 227.56 Cr 24 214.69 Fe 26 11.98 Co 27 628.79 Ni 28 114057.11 Zr 40 1214.93 Nb 41 36.53 Mo 42 7.3 Calculation of Decay Heat Using ORIGEN2.1 The fuel batches operating in TMl-1 Cycle 22 that will be evaluated for decay heat loads are listed in the following table. The table also includes additional information required for simulating the fuels' irradiation using ORIGEN2:

Since the purpose of this calculation is to determine the decay heat load for the highest heat load assembly, the basis of the calculation is 1 fuel assembly.

As noted in Assumption 4.2, the actual enrichments from Reference 3.4 are used to minimize enrichments and maximize decay heat.

Average burnups are taken from Tables 38 and 39 above, with the corresponding Cycle #'s and EFPD for the cycles listed. Per Section 7.1, Batch 24A burnup is taken from core location L-11 in Figure 34.

The fuel assembly's specific power for each cycle is calculated by dividing the burnup by the EFPD, and multiplying by the MTU loading

The MTU loading for a fuel assembly in each batch is taken from Table 14 above.

The grams of U235, U238, and 0 from U02 for each batch are determined as follows:

U235 wt (gms) = Batch loading* (Avg. Enr./100)

106 U238 wt (gms) = Batch loading * (1 - Avg . Enr./100)

106 Oxygen wt (gms) = (U235 wt + U238 wt)/238*2*15.9994

The grams for structural materials are taken from Section 7.2

Exelon>M DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos. , Sheet 202 47 of66 Avg. Bumup Avg. Enr. per Cycle Cycle of Cycle Power Loading U235 wt. U238wt. Oxygen wt.

Batch #ofFA lw/o U235) IMW d/m!U\ Ooeration EFPD IMW\ IMTU\ llmts) (alllS) (l!lllS)

... 19E3 I 4.689 23522 17 700 16.4 0.48918 22937.65 466242.35 65769.63 22570 18 628 17.6 6562 22 720 4.5 52654 22AI I 4.261 27343 20 690 19.2 0.48575 20697.81 465052.19 65308.48 23912 22 720 16.I 5/255 2282 I 4.427 25806 20 690 18.2 0.48607 2151 8.32 464551.68 65351.50 8421 21 647 6.3 13021 22 720 8.8 47248 220 1 4.641 24541 20 690 17.3 0.48777 22637.41 465132.59 65580.06 21524 21 647 16.2 7929 22 720 5.4 53994 23A2 I 3.945 24241 21 647 18.3 0.48843 19268.56 469161.44 65668.80 24263 22 720 16.5 48504 24A I 4.077 28667 22 720 19.3 0.48575 19804.03 465945.97 65308.48 28667 The decay heat loads are calculated using ORIGEN2 and the batch-specific parameters from the table above. The input decks (identifiable by batch number) and a typical batch file , dto.bat, are listed in Appendix A.

The ORIGEN2 input deck is set up to deplete each fuel batch for the requisite number of cycles, and then to decay the isotopic mix after Cycle 22 shutdown. The ORIGEN2 decay heat results are shown in the following table, with decay times in days.

Exelon>M DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 I System Nos.

202 I 48 Sheet of66 Decay Heat (Watts) for Various Decay Times after TMl-1 Cycle 22 Shutdown Batch T=O 30.0D 90.0D 180.0D 365.0D 548.0D 730.0D 913.0D 1095.0D B19E 2.73E+05 1.04E+04 7.18E+03 5.43E+03 3.80E+03 2.99E+03 2.47E+03 2.10E+03 1.85E+03 B22A 9.63E+o5 2.59E+04 1.55E+04 1.0SE+04 6.44Ei-03 4.70Ei-03 3.65Ei-03 2.94Ei-03 2.44Ei-03 B22B 5.28E+05 1.53E+04 9.46E+03 6.54E+03 4.16E+03 3.11E+03 2.48E+03 2.06E+03 1.76E+03 B22D 3.26E+05 1.17E+04 7.95E+03 5.93E+03 4.09E+03 3.20E+03 2.63E+03 2.23E+03 1.95E+03 B23A 9.82E+05 2.61E+04 1.56E+04 1.0SEi-04 6.38E+03 4.63E+03 3.58E+03 2.86E+03 2.37E+03 B24A 1.16Ei-06 2.62Ei-04 1.42E+04 8.54E+03 4.53E+03 3.07E+03 2.26E+03 1.72E+03 1.37E+03 Max "1.16E+06 ,.. 2.62E+04 "1.56E+04 "1.05E+04 ,.. 6.44E+03 ,.. 4.70E+03 ,.. 3.65E+03 ,.. 2.94E+03 ,.. 2.44E+03 For decay times of 180 days and beyond, Batch 22A yields the highest decay heat loads, which will therefore be the basis for the downstream zirc fire analysis.

The Batch 22A results from 365 to 1095 days of decay were plotted in Excel and fit with a 3rd order polynomial trendline (see figure below).

ORIGEN2 Polynomial Fit Difference Decay Heat (Watts) Decay Time (days) Decay Time (days) %

6.44E+o3 365 364.9 0.0 4.70E+o3 548 549.0 0.2 3.65E+o3 730 727.0 -0.4 2.94E+03 913 916.4 0.4 2.44E+03 1095 1093.6 -0.1 The following polynomial fit represents this data for decay heat (x) between 2.44 and 6.44 kW. As seen in the table above, the polynomial fit is in good agreement with the individual ORIGEN2 data points. A 1% multiplier is applied to the Excel fit for conservatism.

Y (days)= 1.01*[*9.22479E-09*x3 +1.59163E-04*x2

1.01332E+OO*x + 2.75333E+03]

Exelon.M DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 I Rev. No.

1 I System Nos.

202 I Sheet 49of66 Maximum Fuel Assembly Decay Heat (Watts) for Various Decay Times after TMl-1 Cycle 22 Shutdown 1,100 y = -9.22479E-09x 3 +1.59163E-04x 2 -1.01332E+00x + 2.75333E+03 1,000 900 V>

> 800 ro

~

QI E 700 I=

ro u 600 QI 0

500 400 300 2.0E+03 2.SE+03 3.0E+03 3.SE+03 4.0E+03 4.SE+03 5.0E+03 5.SE+03 6.0E+03 6.SE+03 7.0E+03 Decay Heat (Watts)

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Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No. , System Nos. , Sheet 1 202 50of66 APPENDIX A ORIGEN2 Input Decks and Typical Job Batch File

Exelon . DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos. , Sheet 202 51 of66 dto-b19e.inp

-1

-1 BAS Grams of Heavy Metal for 1 Mk-B-HTP Fuel Assembly RDA PLACE FUEL into vectors -1 LIP 0 0 0 LIB 0 1 2 3 604 605 606 9 3 0 1 39 PHO 0 0 0 10 RDA READ FUEL COMPOSITION INP -1 1 1 1 1 RDA TIT IRRADIATION OF TMI Batch 19E Fuel to EOC-22 MOV -1 1 0 1.0 HED 1 CHARGE RDA BURN TO EOC-22 BUP IRP 50.0 16.4 1 2 4 2 IRP 100.0 16.4 2 3 4 0 IRP 150.0 16.4 3 4 4 0 IRP 200.0 16.4 4 5 4 0 IRP 250.0 16.4 5 2 4 0 IRP 300.0 16.4 2 3 4 0 IRP 350.0 16.4 3 4 4 0 IRP 400.0 16.4 4 5 4 0 IRP 450.0 16.4 5 6 4 0 IRP 500.0 16.4 6 3 4 0 IRP 550.0 16.4 3 4 4 0 IRP 600.0 16.4 4 5 4 0 IRP 650.0 16.4 5 6 4 0 IRP 700.0 16.4 6 7 4 0 IRP 750.0 17.6 7 4 4 0 IRP 800.0 17.6 4 5 4 0 IRP 850.0 17.6 5 6 4 0 IRP 900.0 17.6 6 7 4 0 IRP 950.0 17.6 7 8 4 0 IRP 1000.0 17.6 8 5 4 0 IRP 1050.0 17.6 5 6 4 0 IRP 1100.0 17.6 6 7 4 0 IRP 1150.0 17.6 7 8 4 0 IRP 1200.0 17.6 8 9 4 0 IRP 1250.0 17.6 9 6 4 0 IRP 1300.0 17.6 6 7 4 0 IRP 1328.0 17.6 7 8 4 0 IRP 1378.0 4.5 8 9 4 0 IRP 1428.0 4.5 9 10 4 0 IRP 1478.0 4.5 10 7 4 0 IRP 1528.0 4.5 7 8 4 0 IRP 1578.0 4.5 8 9 4 0 IRP 1628. 0 4.5 9 10 4 0 IRP 1678.0 4.5 10 11 4 0

Exelon.M DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos.

202 I Sheet 52of66 IRP 1728.0 4.5 11 8 4 0 IRP 1778.0 4.5 8 9 4 0 IRP 1828.0 4.5 9 10 4 0 IRP 1878.0 4.5 10 11 4 0 IRP 1928.0 4.5 11 12 4 0 IRP 1978.0 4.5 12 9 4 0 IRP 2013.0 4.5 9 10 4 0 IRP 2048.0 4.5 10 11 4 0 BUP OPTL 4*8 5 8 8 8 7 15*8 OPTA 4*8 5 8 5 8 7 15*8 OPTF 4*8 5 8 5 8 7 15*8 OUT -11 1 -1 0 MOV 11 1 0 1.0 MOVE EOC22 To Vector 1 HED 1 CHARGE RDA Decay of Discharge Fuel DEC 30.0 1 2 4 2 DEC 90.0 2 3 4 0 DEC 180.0 3 4 4 0 DEC 365.0 4 5 4 0 DEC 548.0 5 6 4 0 DEC 730.0 6 7 4 0 DEC 913.0 7 8 4 0 DEC 1095.0 8 9 4 0 HED 1 T=O OUT -9 1 -1 0 END 2 922350 22937.65 922380 466242.35 0 0.0 U02 4 080000 65913.82 130000 5.99 220000 10.78 240000 227.56 Fuel 4 260000 214.69 270000 11.98 280000 628.79 400000 114057.11 Fuel 4 410000 1214.93 420000 36.53 0 0.0 Fuel 0

END dto-b22a.inp

-1

-1 BAS Grams of Heavy Metal for 1 Mk-B-HTP Fuel Assembly RDA PLACE FUEL into vectors -1 LIP 0 0 0 LIB 0 1 2 3 604 605 606 9 3 0 1 39 PHO 0 0 0 10 RDA READ FUEL COMPOSITION INP -1 1 1 1 1 RDA TIT IRRADIATION OF TMI Batch 22A Fuel to EOC-22 MOV -1 1 0 1.0 HED 1 CHARGE RDA BURN TO EOC-22 BUP

Exelon.M DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos.

202 I Sheet 53 of66 IRP 50.0 19.2 1 2 4 2 IRP 100.0 19.2 2 3 4 0 IRP 150.0 19.2 3 4 4 0 IRP 200.0 19.2 4 5 4 0 IRP 250.0 19.2 5 2 4 0 IRP 300.0 19.2 2 3 4 0 IRP 350.0 19.2 3 4 4 0 IRP 400.0 19.2 4 5 4 0 IRP 450.0 19.2 5 6 4 0 IRP 500.0 19.2 6 3 4 0 IRP 550.0 19.2 3 4 4 0 IRP 600.0 19.2 4 5 4 0 IRP 650.0 19.2 5 6 4 0 IRP 690.0 19.2 6 7 4 0 IRP 740.0 16.1 7 4 4 0 IRP 790.0 16.1 4 5 4 0 IRP 840.0 16.1 5 6 4 0 IRP 890.0 16.1 6 7 4 0 IRP 940.0 16.1 7 8 4 0 IRP 990.0 16.1 8 5 4 0 IRP 1040.0 16.1 5 6 4 0 IRP 1090.0 16.1 6 7 4 0 IRP 1140.0 16.1 7 8 4 0 IRP 1190.0 16.1 8 9 4 0 IRP 1240.0 16.1 9 6 4 0 IRP 1290.0 16.1 6 7 4 0 IRP 1340.0 16.1 7 8 4 0 IRP 1375.0 16.1 8 9 4 0 IRP 1410.0 16.1 9 10 4 0 BUP OPTL 4*8 5 8 8 8 7 15*8 OPTA 4*8 5 8 5 8 7 15*8 OPTF 4*8 5 8 5 8 7 15*8 OUT -10 1 -1 0 MOV 10 1 0 1.0 MOVE EOC22 To Vector 1 HED 1 CHARGE RDA Decay of Discharge Fuel DEC 30.0 1 2 4 2 DEC 90.0 2 3 4 0 DEC 180.0 3 4 4 0 DEC 365.0 4 5 4 0 DEC 548.0 5 6 4 0 DEC 730.0 6 7 4 0 DEC 913.0 7 8 4 0 DEC 1095.0 8 9 4 0 HED 1 T=O OUT -9 1 -1 0 END 2 922350 20697.81 922380 465052.19 0 0.0 U02 4 080000 65452.67 130000 5.99 220000 10.78 240000 227.56 Fuel 4 260000 214.69 270000 11. 98 280000 628.79 400000 114057 .11 Fuel

Exel~n.M DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No. , System Nos. , Sheet 1 202 54 of66 4 410000 1214.93 420000 36.53 0 0.0 Fuel 0

END dto-b22b.inp

-1

-1 BAS Grams of Heavy Metal for 1 Mk-B-HTP Fuel Assembly RDA PLACE FUEL into vectors -1 LIP 0 0 0 LIB 0 1 2 3 604 605 606 9 3 0 1 39 PHO 0 0 0 10 RDA READ FUEL COMPOSITION INP -1 1 1 1 1 RDA TIT IRRADIATION OF TMI Batch 22B Fuel to EOC-22 MOV -1 1 0 1.0 HED 1 CHARGE RDA BURN TO EOC-22 BUP IRP 50.0 18.2 1 2 4 2 IRP 100.0 18.2 2 3 4 0 IRP 150.0 18.2 3 4 4 0 IRP 200.0 18.2 4 5 4 0 IRP 250.0 18.2 5 2 4 0 IRP 300.0 18.2 2 3 4 0 IRP 350.0 18.2 3 4 4 0 IRP 400.0 18.2 4 5 4 0 IRP 450.0 18.2 5 6 4 0 IRP 500.0 18.2 6 3 4 0 IRP 550.0 18.2 3 4 4 0 IRP 600.0 18.2 4 5 4 0 IRP 650.0 18.2 5 6 4 0 IRP 690.0 18.2 6 7 4 0 IRP 740.0 6.3 7 4 4 0 IRP 790.0 6.3 4 5 4 0 IRP 840.0 6.3 5 6 4 0 IRP 890.0 6.3 6 7 4 0 IRP 940.0 6.3 7 8 4 0 IRP 990.0 6.3 8 5 4 0 IRP 1040.0 6.3 5 6 4 0 IRP 1090.0 6.3 6 7 4 0 IRP 1140.0 6.3 7 8 4 0 IRP 1190.0 6.3 8 9 4 0 IRP 1240.0 6.3 9 6 4 0 IRP 1290.0 6.3 6 7 4 0 IRP 1337.0 6.3 7 8 4 0 IRP 1387.0 8.8 8 9 4 0 IRP 1437.0 8.8 9 10 4 0 IRP 1487.0 8.8 10 7 4 0

Exelon.M DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No. , System Nos. ,

1 202 Sheet 55 of 66 IRP 1537.0 8.8 7 8 4 0 IRP 1587.0 8.8 8 9 4 0 IRP 1637.0 8.8 9 10 4 0 IRP 1687.0 8.8 10 11 4 0 IRP 1737.0 8.8 11 8 4 0 IRP 1787.0 8.8 8 9 4 0 IRP 1837.0 8.8 9 10 4 0 IRP 1887.0 8.8 10 11 4 0 IRP 1937.0 8.8 11 12 4 0 IRP 1987.0 8.8 12 9 4 0 IRP 2022.0 8.8 9 10 4 0 IRP 2057.0 8.8 10 11 4 0 BUP OPTL 4*8 5 8 8 8 7 15*8 OPTA 4*8 5 8 5 8 7 15*8 OPTF 4*8 5 8 5 8 7 15*8 OUT -11 1 -1 0 MOV 11 1 0 1.0 MOVE EOC22 To Vector 1 HED 1 CHARGE RDA Decay of Discharge Fuel DEC 30.0 1 2 4 2 DEC 90.0 2 3 4 0 DEC 180.0 3 4 4 0 DEC 365.0 4 5 4 0 DEC 548.0 5 6 4 0 DEC 730.0 6 7 4 0 DEC 913.0 7 8 4 0 DEC 1095.0 8 9 4 0 HED 1 T=O OUT -9 1 -1 0 END 2 922350 21518.32 922380 464551.68 0 0.0 U02 4 080000 65495.69 130000 5.99 220000 10.78 240000 227.56 Fuel 4 260000 214.69 270000 11.98 280000 628.79 400000 114057.11 Fuel 4 410000 1214.93 420000 36.53 0 0.0 Fuel 0

END dto-b22d.inp

-1

-1 BAS Grams of Heavy Metal for 1 Mk-B-HTP Fuel Assembly RDA PLACE FUEL into vectors -1 LIP 0 0 0 LIB 0 1 2 3 604 605 606 9 3 0 1 39 PHO 0 0 0 10 RDA READ FUEL COMPOSITION INP -1 1 1 1 1 RDA TIT IRRADIATION OF TMI Batch 22D Fuel to EOC-22

Exelon>M DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos. , Sheet 202 56 of66 MOV -1 1 0 1. 0 HED 1 CHARGE RDA BURN TO EOC-22 BUP IRP 50.0 17.3 1 2 4 2 IRP 100.0 17 . 3 2 3 4 0 IRP 150.0 17.3 3 4 4 0 IRP 200.0 17.3 4 5 4 0 IRP 250.0 17.3 5 2 4 0 IRP 300.0 17.3 2 3 4 0 IRP 350.0 17.3 3 4 4 0 IRP 400 . 0 17.3 4 5 4 0 IRP 450 . 0 17.3 5 6 4 0 IRP 500 . 0 17.3 6 3 4 0 IRP 550 . 0 17.3 3 4 4 0 IRP 600.0 17 . 3 4 5 4 0 IRP 650.0 17.3 5 6 4 0 IRP 690.0 17.3 6 7 4 0 IRP 740.0 16.2 7 4 4 0 IRP 790 . 0 16.2 4 5 4 0 IRP 840.0 16.2 5 6 4 0 IRP 890.0 16.2 6 7 4 0 IRP 940.0 16.2 7 8 4 0 IRP 990 . 0 16.2 8 5 4 0 IRP 1040.0 16.2 5 6 4 0 IRP 1090.0 16.2 6 7 4 0 IRP 1140.0 16.2 7 8 4 0 IRP 1190.0 16.2 8 9 4 0 IRP 1240 . 0 16.2 9 6 4 0 IRP 1290 . 0 16 . 2 6 7 4 0 IRP 1337.0 16.2 7 8 4 0 IRP 1387.0 5.4 8 9 4 0 IRP 1437 . 0 5.4 9 10 4 0 IRP 1487.0 5.4 10 7 4 0 IRP 1537.0 5.4 7 8 4 0 IRP 1587.0 5.4 8 9 4 0 IRP 1637.0 5.4 9 10 4 0 IRP 1687.0 5.4 10 11 4 0 IRP 1737.0 5.4 11 8 4 0 IRP 1787.0 5.4 8 9 4 0 IRP 1837.0 5.4 9 10 4 0 IRP 1887.0 5.4 10 11 4 0 IRP 1937.0 5.4 11 12 4 0 IRP 1987.0 5.4 12 9 4 0 IRP 2022 . 0 5.4 9 10 4 0 IRP 2057.0 5.4 10 11 4 0 BUP OPTL 4*8 5 8 8 8 7 15*8 OPTA 4*8 5 8 5 8 7 15*8 OPTF 4*8 5 8 5 8 7 15*8 OUT -u 1 -1 0

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Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis 'Rev. No. , System Nos. , Sheet C-1101-202-E410-476 1 202 57 of66 MOV 11 1 0 1.0 MOVE EOC22 To Vector 1 HED 1 CHARGE RDA Decay of Discharge Fuel DEC 30.0 1 2 4 2 DEC 90.0 2 3 4 0 DEC 180.0 3 4 4 0 DEC 365.0 4 5 4 0 DEC 548.0 5 6 4 0 DEC 730.0 6 7 4 0 DEC 913.0 7 8 4 0 DEC 1095.0 8 9 4 0 HED 1 T=O OUT -9 1 -1 0 END 2 922350 22637.41 922380 465132.59 0 0.0 U02 4 080000 65724.25 130000 5.99 220000 10.78 240000 227.56 Fuel 4 260000 214.69 270000 11.98 280000 628.79 400000 114057.11 Fuel 4 410000 1214.93 420000 36.53 0 0.0 Fuel 0

END dto-b23a.inp

-1

-1 BAS Grams of Heavy Metal for 1 Mk-B-HTP Fuel Assembly RDA PLACE FUEL into vectors -1 LIP 0 0 0 LIB 0 1 2 3 604 605 606 9 3 0 1 39 PHO 0 0 0 10 RDA READ FUEL COMPOSITION INP -1 1 1 1 1 RDA TIT IRRADIATION OF TMI Batch 23A Fuel to EOC-22 MOV -1 1 0 1.0 HED 1 CHARGE RDA BURN TO EOC-22 BUP IRP 50.0 18.3 1 2 4 2 IRP 100.0 18.3 2 3 4 0 IRP 150.0 18.3 3 4 4 0 IRP 200.0 18.3 4 5 4 0 IRP 250.0 18.3 5 2 4 0 IRP 300.0 18.3 2 3 4 0 IRP 350.0 18.3 3 4 4 0 IRP 400.0 18.3 4 5 4 0 IRP 450.0 18.3 5 6 4 0 IRP 500.0 18.3 6 3 4 0 IRP 550.0 18.3 3 4 4 0 IRP 600.0 18.3 4 5 4 0 IRP 647.0 18.3 5 6 4 0

Exelon.M DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos.

202 I Sheet 58of66 IRP 697.0 16.5 6 7 4 0 IRP 747.0 16.5 7 4 4 0 IRP 797.0 16.5 4 5 4 0 IRP 847.0 16.5 5 6 4 0 IRP 897.0 16.5 6 7 4 0 IRP 947.0 16.5 7 8 4 0 IRP 997.0 16.5 8 5 4 0 IRP 1047.0 16.5 5 6 4 0 IRP 1097.0 16.5 6 7 4 0 IRP 1147.0 16.5 7 8 4 0 IRP 1197 .0 16.5 8 9 4 0 IRP 1247.0 16.5 9 6 4 0 IRP 1297.0 16.5 6 7 4 0 IRP 1332.0 16.5 7 8 4 0 IRP 1367.0 16.5 8 9 4 0 BUP OPTL 4*8 5 8 8 8 7 15*8 OPTA 4*8 5 8 5 8 7 15*8 OPTF 4*8 5 8 5 8 7 15*8 OUT -9 1 -1 0 MOV 9 1 0 1.0 MOVE EOC22 To Vector 1 HED 1 CHARGE RDA Decay of Discharge Fuel DEC 30.0 1 2 4 2 DEC 90.0 2 3 4 0 DEC 180.0 3 4 4 0 DEC 365.0 4 5 4 0 DEC 548.0 5 6 4 0 DEC 730.0 6 7 4 0 DEC 913.0 7 8 4 0 DEC 1095.0 8 9 4 0 HED 1 T=O OUT -9 1 -1 0 END 2 922350 19268.56 922380 469161.44 0 0.0 U02 4 080000 65812.99 130000 5.99 220000 10.78 240000 227.56 Fuel 4 260000 214.69 270000 11. 98 280000 628.79 400000 114057.11 Fuel 4 410000 1214.93 420000 36.53 0 0.0 Fuel 0

END dto-b24a.inp

-1

-1 BAS Grams of Heavy Metal for 1 Mk-B-HTP Fuel Assembly RDA PLACE FUEL into vectors -1 LIP 0 0 0 LIB 0 1 2 3 604 605 606 9 3 0 1 39 PHO 0 0 0 10

Exelon.y DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos.

202 I 59 Sheet of 66 RDA READ FUEL COMPOSITION INP -1 1 1 1 1 RDA TIT IRRADIATION OF TMI Batch 24A Fuel to EOC-22 MOV -1 1 0 1. 0 HED 1 CHARGE RDA BURN TO EOC-22 BUP IRP 50.0 19.3 1 9 4 2 IRP 100.0 19.3 9 2 4 0 IRP 150.0 19.3 2 9 4 0 IRP 200.0 19.3 9 3 4 0 IRP 250.0 19.3 3 9 4 0 IRP 300.0 19.3 9 4 4 0 IRP 350.0 19.3 4 9 4 0 IRP 400.0 19.3 9 5 4 0 IRP 450.0 19.3 5 9 4 0 IRP 500.0 19.3 9 6 4 0 IRP 550.0 19.3 6 9 4 0 IRP 600.0 19.3 9 7 4 0 IRP 650.0 19.3 7 9 4 0 IRP 685.0 19.3 9 8 4 0 IRP 720.0 19.3 8 9 4 0 BUP OPTL 4*8 5 8 8 8 7 15*8 OPTA 4*8 5 8 5 8 7 15*8 OPTF 4*8 5 8 5 8 7 15*8 OUT -9 1 -1 0 MOV 9 1 0 1.0 MOVE EOC22 To Vector 1 HED 1 CHARGE RDA Decay of Discharge Fuel DEC 30.0 1 2 4 2 DEC 90.0 2 3 4 0 DEC 180.0 3 4 4 0 DEC 365.0 4 5 4 0 DEC 548.0 5 6 4 0 DEC 730.0 6 7 4 0 DEC 913.0 7 8 4 0 DEC 1095.0 8 9 4 0 HED 1 T=O OUT -9 1 -1 0 END 2 922350 19804.03 922380 465945.97 0 0.0 U02 4 080000 65452.67 130000 5.99 220000 10.78 240000 227.56 Fuel 4 260000 214.69 270000 11.98 280000 628.79 400000 114057.11 Fuel 4 410000 1214.93 420000 36.53 0 0.0 Fuel 0

END

Exelon~

DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos.

202 I Sheet 60 of66 dto.bat Note: DOSBox was used to run ORIGEN2 on the nfw-ksq-01 server. D: \dosprogs was mounted as the D: drive when this batch file was run.

echo off echo ***********************************************************************

echo ***********************************************************************

echo ** **

echo ** ORIGEN2 **

echo ** Oak Ridge Isotope GENeration and Depletion Code **

echo ** Version 2.1 (8-1-91) **

echo ** **

echo ***********************************************************************

echo ** **

echo ** Developed by: Oak Ridge National Laboratory **

echo ** Chemical Technology Division **

echo ** **

echo ** Technical

Contact:

Scott B. Ludwig **

echo ** (615) 574-7916 FTS 624-7916 **

echo ** **

echo ** Distributed by: Radiation Shielding Information Center (RSIC) **

echo ** Oak Ridge National Laboratory **

echo ** P.O. Box 2008 **

echo ** Oak Ridge, TN 37831 **

echo ** (615) 574-6176 FTS 624-6176 **

echo ***********************************************************************

pause echo** Execution continuing ... **

echo ***********************************************************************

echo ** **

echo ** Version 2.1 (8-1-91) for mainframes and 80386 or 80486 PCs **

echo ** **

copy dto-b24a.inp tape5.inp >nul REM (NOT USED IN THIS CASE) copy samp 2.u3 tape3.inp >nul copy d:\origen2\libs\decay.lib+d:\orig;"n2\libs\pwrue.lib tape9.inp >nul copy d:\origen2\libs\gxuo2brm.lib tapelO.inp >nul d:\origen2\code\origen2 rem combine and save files from run copy tape12.out+tape6.out dto-b24a.u6 >nul copy tape13.out+tapell.out dto-b24a.out >nul ren tape7.out dto-b24a.pch ren tape15.out dto-b24a.dbg ren tape16.out dto-b24a.vxs ren tape50.out dto-b24a.ech rem cleanup files del tape*.inp del tape*.out echo ***********************************************************************

Exelon WO DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis l Rev. No.

C-1101-202-E410-476 I 1 ISystem Nos.

202 I Sheet 61 of66 echo ******************* 0 R I GE N 2 - Version 2.1 ***********************

echo *********************** Execution Completed ***************************

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Exelon>M DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 I System Nos.

202 I Sheet 62 of66 APPENDIXB Alloy 718 Chemistry

CfilJt' B637 - 16 TABLE 1 Chemical Requirements Bement Compos Ilion LIT111S, %

UNS UNS UNS UNS UNS UNS UNS N07252 N07001 N07500 UNS N07750 N07718 N07022 N07208 {Formelly (Formerly IFormeJty N07740 iFormeJty fFormelly Grade 6891 Grade685) Grade684) Grade688} Grade Camon O.OIOmax OD4--0.08 0. 10-0.20 O.lll-0.10 0.15max 0.005--0.08 0.08max 0.08 max Manganese 0.5max 03max 0.50max 1.00 max 0.7Smax l.OOmax l.OOmax 0.35max Siiccn 0.08max O.IS max 0.50max 0.75 max 0.75 max l.OOmax 0.50max 0.35max Pn~horl.IS 0.025max O.OIS max O.OISmax 0.030 max O.Ol5max 0.030 max o.015max Sulfur 0.015max OD15max 0.015max 0.030 max 0.015max 0.030 max 0.01 max 0.015 max Chromiim 20.0-21.4 18.5-20.5 18.00-20.00 18.00-21.00 15.00-20 .00 23.50-25.50 14.00-17.00 17.0-21.0 Cooall l.Omax 9.0-11.0 9.00-11.00 12.00-15.00 13.00-20 .00 IS .00-22.00 l.OOmax .. l.Omax ..

Molybden t.m IS.5-17.4 8.0-9.0 9.00-10.50 3.SO-S.00 3.00..S.OO 2.00max 2.8<>-3.30 Cdt.mllium 0.70-1.20 4.7&--5.50 (Nbl ..- tantalLITI Tllanit.m 1.90-2.30 2.25-2.75 2.75-325 2.5()..3.25 O.S0-2.50 2.25-2.75 0.65-1.15 AllninLITI 0.5max 138-1.65 0.75-1.25 120-1.60 2. 5()..3.25 0.20-2.00 0.40-1.00 020-0.80 Zircoriim O!l20 max O!l2-0.12 Boron 0.006max 0.003--0.010 0.003-0DI O.ooJ--0.01 0.003--0.01 0.0008--0.006 0.006 max Iron 1.smax 1.5 max S.OOmax 2.00 max 4.00max 3.00max S.00-9.00 remard!!r 0 Copper O.Smax 0.1 max 0.50 max O.ISmax 0.50 maic O.SOmax O.JOmax Nicl!BI remarder" rem.-.der" remM'lder" remainder" remoWlder" remairoe.r" 70.00,,.., 50.0-55.0 Tantaltm 0.2max 0.1 max Cdimbum 0.50-2.50 02 max (lliobi.m)

Ti.rgs1en 0.8max 0.5 max UNS N07060 fFormelly UNS N07752 UNSN09925 UNS Nom5 GradeOOA Carbon O.IOmax 0 ,020--0.060 0.03max O.oJ max l.langanese 1.oomax 1.00 max l.Omax 035 max Silllalr\ l.OOmax 0.50 max O.Smax 020 max Ptl~horl.IS 0.008 max 0.03max 0.015 max Sulfur O.OISmax 0.003 max 0.03max 0.010 max Chromttm 18.00-21.00 14.50-17.00 19.5-22.5 19.00-22.50 Cobalt O.OSO max Molybdent.m 2.5-3.5 7.00...9.50 Cdtinbum 0.70-1.20 0.5 max fNb aiy} 2 .75-4.00

{Nb} + tantall.ITI Tlllntt.m 1.80-2.70 2.25-2.75 1.9-2.40 t.00-1.70 AILllWILITI O.S0-1.80 0.40-1.00 0.1--0.5 035 max Boron 0.007 fflAJ(

Iron 3.00max 5.00-9.00 22.0 m., remarder" Copper 0.50 max 1.5-3.0 Ztrcawtin o.oso max Vanadlt.m 0.10 max Nicl!el remarder" 70.0 rnn 42.<>-46.0 55.0-59.0

.. 11 determined.

"The dement shall be de1e llT'Sled anttmeticaly by ddlerooce.

Exelon~

DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 ISystem Nos.

202 I Sheet 63 of66 ATTACHMENT 2 Specific Heat Study A study was performed to determine a suitable temperature to be used to calculate the specific heats for the materials. Each material's specific heat is governed by an equation (or set of equations) provided in References 3.17, 3.18, and 3.19 (shown below):

Where:

Cp. uo2 = specific heat of U02, J/g-K e =535.85 K Eo = 157.7707*103 J/mol K1 = 80.1314 J/mol-K K2 = 32.845*10 4 J/mol-K2 KJ = 23.62183*106 J/mol R = 8.134 J/mol-K (ideal gas constant)

Tk =Temperature, K Auo2 = atomic weight of uranium dioxide, g/mole Cp,Ms = 0.2375 + 0.0001591

T for 273K 5 T 5 llOOK Cp,Ms = -13.4034 + 0.01256

T for llOOK 5 T 5 1140K cp,Ms = 6.9160 - 0.005264

T for 1140K 5 T 5 1250K cp,Ms = 0.2141 + 0.0000975

T for 1250K 5 T 5 1600K Where:

Cp, Ms = specific heat of M5, J/g-K T =Temperature, K cp,ss = (0.1122 + 3.222x10 -s

T) * (4.184 cal J) 1 Where Cp,ss =specific heat of CF3 stainless steel, J/g-K T =Temperature, K lnconel Alloy 718 specific heats are provided in Appendix 7.5.

Exelon~

DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 I Rev. No. , System Nos. , Sheet 1 202 64 of66 In order to determine how these parameters affect the heat-up time for the fuel assembly, the heat transfer equation (Equation 6.5.10-2) was solved for time:

l\T t = 4 * (muoz

Cp,UOz +(PMS* VMS.Rods+ mMs,Grids)

Cp,MS + mss

Cp,SS + mHMP

Cp,HMP)

For simplicity, the values calculated for Lff, q, muo2, PMS* VMs, mMs,Gnds, mss, and m1ncone1 in section 6.5 of this calculation are used and assumed to be constant over the temperature range.

Lff = 857 K q = 5270W muo2 =5.51 Es g PMs = 6.50 g/cm 3

=

VMs,Rods 18029.6 cm 3 mMS,Grids = 7620.9 9 mss = 16692.2 g mHMP =1197.7 9

Exelonuo DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 IRev. No.

1 I System Nos. , Sheet 202 65 of66 The following table calculates the heat-up time for the fuel assembly using the temperature dependent specific heat values.

Temperature U02 MS CF3SS Alloy 718 Heat-up (K) (J/g-K) (J/g-K) (J/g-K) (J/g-K) Time (hrs) 300 0.236 0.285 0.510 0.451 7.90 316 0.242 0.288 0.512 0.459 8.07 400 0.266 0.301 0.523 0.488 8.74 500 0.282 0.317 0.537 0.522 9.25 533 0.286 0.322 0.541 0.528 9.38 600 0.292 0.333 0.550 0.546 9.60 700 0.300 0.349 0.564 0.574 9.89 745 0.302 0.356 0.570 0.583 10.00 800 0.305 0.365 0.577 0.597 10.13 900 0.310 0.381 0.591 0.675 10.35 1000 0.314 0.397 0.604 0.708 10.55 1100 0.318 0.413 0.618 0.693 10.74 1140 0.319 0.915 0.623 0.687 13.62 1173 0.320 0.741 0.627 0.682 12.67 1200 0.321 0.599 0.631 0.678 11.89 1250 0.323 0.336 0.638 0.671 10.45 1300 0.325 0.341 0.645 0.663 10.53 Note: The specific heat values for lnconel Alloy 718 above 1088 Kare extrapolated based on the last two values in the Appendix 7.5 table.

Exelon WO DESIGN ANALYSIS SHEET Nuclear

Subject:

DECOM Spent Fuel Pool TH Analysis I Design Analysis C-1101-202-E410-476 I Rev. No.

1 I System Nos.

202 I Sheet 66of66 The values from the table are plotted in the figure below.

Zr Fire Heat-up Time vs Temperature 16 1-!X>O 0 900 14 0 .800 12 0 700 10

~

. a-

! 8 0.500 * ~

!... I I

..,:i' 1;i 0400 :

I I

I

- -+-- - -.-- o . ~oo Midpoint Temperature a.zoo 881F /745 K 2

0 !.00 Initial Temperature Final Temperature 110 F / 316 K 900C/1173 K 0 0.000 0 zoo 600 800 1000 1200 1AOO nmper.iture [Kl

- - H*at-upT1me(hts1 _._ uo2 - - M5 ---- cB 55 (J/g*K) All 0"(71 B (J /g- K)

IJ/g*KI (J/g*KI The figure clearly shows that the heat-up time increases with increasing material temperature for all materials except for lnconel Alloy 718. However, even though the specific heat for lnconel decreases at the higher end of the temperature band, it does not have a significant impact on the heat-up time due to the minimal amount of mass it is applying to in the calculation (0.8% of total mass). Overall, this means that as the materials in the fuel assembly reach higher temperatures, they would heat up more slowly.

Therefore, using a temperature at or below the midpoint of the temperature range would be conservative with respect to the assembly heat-up. For the purposes of this analysis, the specific heats were calculated at the midpoint temperature of 881°F (745K).

v t e Letter Sequence Request
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ML19182A104

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2.0 BACKGROUND

5.0 TECHNICAL EVALUATION

6.0 CONCLUSION

10.0 REFERENCES

2.0 BACKGROUND

5.0 TECHNICAL EVALUATION

6.0 CONCLUSION

10.0 REFERENCES

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