ML23033A103

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Supplement to License Amendment Request - Proposed Changes to TMl-2 Possession Only License and Technical Specifications
ML23033A103
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 01/27/2023
From: Hazelhoff A
EnergySolutions, TMI-2 Solutions
To:
Office of Nuclear Material Safety and Safeguards, Document Control Desk
Shared Package
ML23033A153 List:
References
TMl2-RA-COR-2023-0002
Download: ML23033A103 (1)
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Text

{{#Wiki_filter:Attachment 2 contains Proprietary Information. Withhold from public disclosure in accordance with 10 CFR 2.390. When Attachment 2 is removed, this document is Decontrolled. 

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  • TMl-2 SOLUTIONS January 27, 2023 TMl2-RA-COR-2023-0002 10 CFR 50.90 10 CFR 50.91 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 2 (TMl-2)

NRC Possession Only License No. DPR 73 NRC Docket No. 50-320

Subject:

Supplement to License Amendment Request - Proposed Changes to TMl-2 Possession Only License and Technical Specifications

References:

(1) Letter TMl2-RA-COR-2021-0002 from van Noordennen (TMl-2 Solutions, LLC) "License Amendment Request- Three Mile Island, Unit 2, Decommissioning Technical Specifications," (ML21057A046) dated February 19, 2021. (2) Letter TMl2-RA-COR-2021-0010 from van Noordennen (TMl-2 Solutions, LLC) "Supplemental Information to License Amendment Request - Three Mile Island, Unit 2, Decommissioning Technical Specifications," (ML21133A264) dated May, 5, 2021. (3) Letter TMl2-RA-COR-2022-0002 from van Noordennen (TMl-2 Solutions, LLC) "License Amendment Request - Three Mile Island, Unit 2, Decommissioning Technical Specifications, Supplemental Information," (ML22013A177) dated January 7, 2022. (4) Letter TMl2-RA-COR-2022-0007 from van Noordennen (TMl-2 Solutions, LLC) "License Amendment Request - Three Mile Island, Unit 2, Decommissioning Technical Specifications, Response to Questions," dated April 8, 2022. (5) Letter TMl2-RA-COR-2022-0008 from van Noordennen (TMl-2 Solutions, LLC) "Supplemental Information to License Amendment Request- Three Mile Island, Unit 2, Decommissioning Technical Specifications," (ML22101A077) dated April 7, 2022. (6) Letter TMl2-RA-COR-2022-0013 from Lackey (TMl-2 Solutions LLC) 

              "License Amendment Request -                 Three Mile Island, Unit 2, Attachment 2 contains Proprietary Information. Withhold from public disclosure in accordance with 10 CFR 2.390. When Attachment 2 is removed, this document is Decontrolled.

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                                                                                              \S Attachment 2 contains Proprietary Information. Withhold from public disclosure in accordance with 10 CFR 2.390. When Attachment 2 is removed, this document is Decontrolled.

Decommissioning Technical Specifications, Response to Questions," (ML22138A285) dated May 16, 2022. (7) Letter TMl2-RA-COR-2022-0019 from Lackey (TMl-2 Solutions, LLC) 

                "License Amendment Request -                Three Mile Island, Unit 2, Decommissioning Technical Specifications, Response to Requests for Additional Information," (ML22276A024) dated September 29, 2022.

(8) Letter TMl2-RA-COR-2022-0021 from Lackey (TMl-2 Solutions, LLC) 

                "License Amendment Request -                Three Mile Island, Unit 2, Decommissioning Technical Specifications, Response to Requests for Additional Information," (ML22307A082) dated October 31, 2022.

(9) Letter TMl2-RA-COR-2022-0023 from Lackey (TMl-2 Solutions, LLC) 

                "License Amendment Request -                Three Mile Island, Unit 2, Decommissioning Technical Specifications, Response to Request for Information Regarding Radiation Protection Program Ventilation Controls,"

(ML22313A050) dated November 7, 2022. (10) Letter TMl2-RA-COR-2023-0001 from Hazelhoff (TMl-2 Solutions, LLC) 

                "License Amendment Request - Three Mile Island, Unit 2, Decommissioning Technical Specifications, Response to Request for Additional Information Regarding Historical and Cultural Resources," dated January 20, 2023.

Three Mile Island Unit 2 (TMl-2) submitted a License Amendment Request (LAR) (Reference 1) to revise the TMl-2 Possession Only License (POL) to support the transition of TMl-2 from a Post-Defueling Monitored Storage (PDMS) condition to that of a facility undergoing radiological decommissioning (DEGON). References 2 through 10 provided supplemental information and responses to Requests for Additional Information (RAls) to Reference 1. The proposed changes would revise certain requirements contained within the POL and Technical Specifications (TS) and remove the requirements that would no longer be applicable. The proposed changes to the POL and TS are in accordance with 10 CFR 50.36(c)(1) through 10 CFR 50.36(c)(5). The current TS contain Limiting Conditions for PDMS that provide for appropriate functional capability of equipment required for safe storage of the facility. However, the majority of the existing TS are only applicable with the facility in the PDMS condition. Technical Specifications, Limiting Conditions for PDMS and Surveillance Requirements (SRs) that do not apply during decommissioning are being proposed for deletion. The remaining portions of the TS will be renamed the Decommissioning Technical Specifications (DTS). The DTS provide a continuing acceptable level of safety for the facility as it undergoes decommissioning. Attachment 2 contains Proprietary Information. Withhold from public disclosure in accordance with 10 CFR 2.390. When Attachment 2 is removed, this document is Decontrolled.

Attachment 2 contains Proprietary Information. Withhold from public disclosure in accordance with 10 CFR 2.390. When Attachment 2 is removed, this document is Decontrolled. The proposed changes to the TMl-2 Solutions POL and TS also involve relocating administrative controls from Section 6, "Administrative Controls," to the Decommissioning Quality Assurance Plan (DQAP), and subsequently controlling them in accordance with 10 CFR 50.54(a). This relocation is being proposed pursuant to the criteria contained in 10 CFR 50.36 and in accordance with the recommendations, guidance, and purpose of NRG Administrative Letter 95-06, "Relocation of Technical Specification Administrative Controls Related to Quality Assurance." Relocated content will be incorporated into the DQAP verbatim, except for TS sectional cross references, as required by Specification 6.9.2.e. This LAR supplement provides the following:

  • Attachment 1 provides a Response to Request for Additional Information for Reques~ed Licensing Action Regarding Decommissioning Technical Specifications EPID: L-2021-LLA-0038 (RAI 3).
  • Attachment 2 provides the Calculation No. 164090-EN-CAL-004 "Source Term Limitations and Administrative Controls for the TMl-2 Decommissioning Emergency Plan Action Levels," (Proprietary).
  • Attachment 3 provides a Proprietary Information Affidavit requesting that the analysis provided in Attachment 2 be withheld from public disclosure in accordance with 10 CFR 2.390.
  • Attachment 4 contains revised responses to RAls 10 - 13. Change bars indicate revisions to the original responses. These responses supersede those provided in Reference 8.
  • Attachment 5 provides a revised evaluation of certain proposed changes to the TMl-2 POL and TS that have been affected by References 2-10. This Attachment includes a revised No Significant Hazards Consideration (NSHC), which supersedes the NSHC provided in Reference 3. Change bars indicate revisions to the original NSHC. The NSHC provided in Reference 3 remains bounding.
  • Attachment 6 provides a mark-up of the current POL and TS pages, including the TS Bases. This Attachment supersedes the marked-up pages provided in Reference 1.
  • Attachment 7 contains the clean POL and TS pages. This Attachment supersedes the clean POL and TS pages provided in Reference 1.
  • Attachment 8 provides the Regulatory Commitments contained in this submittal.

In accordance with 10 CFR 50.91 (b )( 1), a copy of this submittal has been sent to the Commonwealth of Pennsylvania. Attachment 2 contains Proprietary Information. Withhold from public disclosure in accordance with 10 CFR 2.390. When Attachment 2 is removed, this document is Decontrolled.

Attachment 2 contains Proprietary Information. Withhold from public disclosure in accordance with 10 CFR 2.390. When Attachment 2 is removed, this document is Decontrolled. If there are questions with respect to the content of this document, please contact the TMl-2 Solutions Licensing Manager, Mr. Timothy Devik at 603-384-0239, or by email at trdevik@energysolutions.com. I declare under penalty of perjury that the foregoing is true and correct. Executed on January 27, 2023. Sincerely, CDigitally signed by Amy C Amy C Haze Ihoff J-!~~~ 

               ,\Hazelhoff 0
                      ~023.01 .21 , 6:so,30 Amy C. Hazelhoff Vice President Regulatory Affairs EnergySo/utions, LLC Attachments:
1. Response to Request for Additional Information for Requested Licensing Action Regarding Decommissioning Technical Specifications (RAI 3)
2. Calculation No. 164090-EN-CAL-004, "Source Term Limitations and Administrative Controls for the TMl-2 Decommissioning Emergency Plan Action Levels," (Proprietary)
3. Proprietary Information Affidavit
4. Revision to RAI 10-13 Responses
5. Evaluation of the proposed changes to the TMl-2 POL and TS and Revised NSHC
6. POL, TS and Bases Mark-up Pages
7. POL and TS Clean Pages
8. List of Regulatory Commitments cc: w/Attachments Regional Administrator - NRC Region I NRC Lead Inspector - Three Mile Island Nuclear Station - Unit 2 NRC Project Manager - Three Mile Island Nuclear Station - Unit 2 Attachment 2 contains Proprietary Information. Withhold from public disclosure in accordance with 10 CFR 2.390. When Attachment 2 is removed, this document is Decontrolled.

Attachment 2 contains Proprietary Information. Withhold from public disclosure in accordance with 10 CFR 2.390. When Attachment 2 is removed, this document is Decontrolled. TMl-2 Service List Ken Robuck Director, Bureau of Radiation Protection, President and CEO Department of Environmental Protection, Energy Solutions Commonwealth of Pennsylvania 299 South Main Street, Suite 1700 Rachel Carson State Office BLDG. Salt Lake City, UT 84111 13TH Floor P.O. Box 8469 John Sauger Harrisburg, PA 17105-8469 President and Chief Nuclear Officer Reactor D&D Chief, Division of Nuclear Safety, Bureau of Energy Solutions Radiation Protection, Department of Environmental Protection, 121 W. Trade Street, Suite 2700 Commonwealth of Pennsylvania Charlotte, NC 28202 Rachael Carson State Office BLDG. 13TH Floor Sam Bambino P.O. BOX 8469 Senior Vice President Harrisburg, PA 17105-8469 D&D Operations Energy Solutions Chairman, Board of County Commissioners, 121 W. Trade Street, Suite 2700 Dauphin County Charlotte, NC 28202 112 Market Street 7th Floor Frank Helin Harrisburg, PA 17101 Project Director TMl-2 Solutions Trevor Orth 121 W. Trade Street, Suite 2700 Site Decommissioning Director Charlotte, NC 28202 Three Mile Island Generating Station Route 441 South Russ Workman Middletown, PA 17057 General Counsel Energy Solutions Craig Smith 299 South Main Street, Suite 1700

  • Site Decommissioning Regulatory Assurance Salt Lake City, UT 84111 Lead Three Mile Island Generation Station Daniel F. Stenger Route 441 South Hogan Lovells US LLP Middletown, PA 17057 555 13th St NW Washington, D.C. 20004 Tim Devik TMl-2 Licensing Manager Amy C. Hazelhoff Three Mile Island Generating Station Vice President, Regulatory Affairs Route 441 South Energy Solutions Middletown, PA 17057 299 South Main Street, Suite 1700 Salt Lake City, UT 84111 Attachment 2 contains Proprietary Information. Withhold from public disclosure in accordance with 10 CFR 2.390. When Attachment 2 is removed, this document is Decontrolled.

Attachment 1 TMl2-RA-COR-2023-0002 ATTACHMENT 1 Response to Request for Additional Information for Requested Licensing Action Regarding Decommissioning Technical Specifications (RAI 3) 6 Pages Follow

Attachment 1 TM 12-RA-COR-2023-0002 THREE MILE ISLAND, UNIT No. 2 - REQUEST FOR ADDITIONAL INFORMATION FOR REQUESTED LICENSING ACTION REGARDING DECOMMISSIONING TECHNICAL SPECIFICATIONS EPID: L-2021-LLA-0038 ACCIDENT ANALYSIS: By letter dated February 19, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21057A046), TMl-2 Solutions, LLC (TMl-2 Solutions or licensee) submitted a License Amendment Request (LAR) to remove certain requirements from the TMl-2 Technical Specifications (TS) that restrict activities in the TMl-2 Reactor Building during Post-Defueling Monitored Storage (PDMS). The licensee would like to progress to actively decommissioning the remaining structures, systems, and components that were contaminated in the 1979 accident. Previously, the licensee had analyzed the impacts of a fire in a High Integrity Container (HIC) containing spent ion exchange resins. Subsequently, the licensee determined that the HIC fire scenario was not representative of the activities that would be occurring during decommissioning and submitted supplemental information on January 7, 2022 (ML22013A177). The U.S. Nuclear Regulatory Commission (NRG) staff provided preliminary questions on the information on February 7, 2022 (ML22038A936). The licensee provided a response on April 7, 2022 (ML22101A077), including references and additional analyses on May 8, 2022 (ML22138A302). This request for additional information (RAI) is in response to the latest information provided. Fire is one of the largest risks at a nuclear facility (U.S. Department of Energy (DOE), 1994 ). Fire risk is a product of the likelihood of a fire occurring and the consequences if a fire were to occur. Though minor in impact, fires have occurred at nuclear reactors undergoing decommissioning (e.g., Crystal River, Ft. Calhoun, Indian Point). By the introduction of fuel and energy sources combined with the diverse activities that are necessary to complete decommissioning, the frequency of occurrence of fires has been higher during decommissioning than during operations or, in the case of TMl-2, PDMS. When responding to RAls, the licensee may identify alternative approaches such as management controls, procedures, calculations, or conditions that will ensure the impacts from potential fires during decommissioning will meet established criteria for protection of human health. RAI 3 Offsite Dose Calculations Comment: The offsite dose calculations lack transparency and traceability. Basis: Offsite doses resulting from a potential fire were described in TMl2-RA-COR-2022-0007, LAR TMI- 2 "GPU.Nuclear Calculation 4440-7380-90-017, Revision 4, PDMS Safety Analysis Report (SAR) Section 8.2.5 Fire Analysis Source Terms") of 1

Attachment 1 TMl2-RA-COR-2023-0002 ML22138A302 (May 13, 2022, Attachment 2). The licensee described modifications to previous calculations (revision 3) to account for additional decay and ingrowth, the presence of additional loose contamination, and use of updated dose conversion factors (revision 4). These changes were sufficiently described and appropriate. Staff were able to verify the dose conversion factors that were used and most other parameters, as well as the calculated decay and ingrowth. However, the approach taken for the amount of source material (inventory) that is released as a result of the fire was not clear. In revision 3 of the analysis, the amount released was calculated as a product of two factors: the amount of material available and the fraction of available material that was released to the air. The amount offuel elements available (e.g., Pu, Am) was assumed to be 100 percent or a fraction of 1.0. The amount of Cs and Sr available was assumed to be 1 percent or a fraction of 0.01 for a fire in the reactor basement. These were then multiplied by factors of 8 x 10-4 for actinides and apparently 1.5 x 10-3 for Cs and Sr. Staff could only replicate the basement fire dose of 0.889 mrem by using these factors. The impact is the dose for new ARF of 1.5 x 10-4 does not decrease by a factor of 6.67 but instead would be 0.80 mrem for the basement fire (note RAI #1 on the basis for the ARF). Path Forward: Please verify the combined factors of material available and airborne fraction released in revision 3 of the fire analysis source terms and update the revision 4 analyses as appropriate. TMl-2 Solutions RAI 3 Response: As described in Reference 1, the bases of the original airborne source term used in the legacy PDMS analyses and subsequent updates was not able to be located or reproduced. Therefore, TMl-2 Solutions has performed a new fire analysis based on a known contamination source term to be used in determining off-site releases. This analysis, 164090-EN-CAL-004 "Source Term Limitations and Administrative Controls for the TMl-2 Decommissioning Emergency Plan Action Levels," is provided in Attachment

2. Unlike the legacy PDMS analyses which calculated releases due to non-mechanistic fires, this analysis establishes limits on the accumulation of uncontained radioactive materials to ensure the criteria for the Notification of an Unusual Event (NOUE) will not be exceeded in the event of a postulated fire. As described in Attachment 2, the new fire analysis uses decay corrected 1993 10 CFR Part 61 waste stream data for Dry Active Waste (DAW) (Reference 2), removable contamination smears on a Drone used to survey the Reactor Building in June of 2021, as well as recent smears associated with the Reactor Building elevator and the polar crane.

PDMS SAR Table 5.3-5 defines three mixes of radionuclides found at TMl-2. These 2

Attachment 1 TMl2-RA-COR-2023-0002 mixes are named Defueling, Make-Up and Purification (MUP) and Normal1. The decay corrected mixes from the PDMS data set were compared to the more recent samples from 2021 by converting the mixture fractions to relative inhalation doses. This comparison indicates that the MUP and Normal mixes are most representative of the overall radionuclide mixes in the Reactor Building areas and most other areas of TMl-2. When considering offsite radiation exposure, the MUP mix is more conservative than the Normal mix for both bone and TEDE doses and was used to provide a bounding and conservative mix for use in the fire accident analysis. For those few areas where the Defueling mix is the controlling factor (A Spent Fuel Pool, A Reactor Coolant Bleed Tank Cubicle and the Reactor Building 347" Elevation in the vicinity of the A Core Flood Tank), they will be specifically managed to limit potential combustion sources for DAW source terms. The DAW dose rate limits established by the analysis in Attachment 2 applies to open containers without further administrative controls. The movement of containers is not limited, provided the contents are sealed from potential ignition sources (i.e., closed). Containers holding DAW will be closed or sealed from potential ignition sources prior to movement. Based on the accumulation of combustible waste materials being the limiting factor, cut off dose rates were calculated using MicroShield version 8.03 for uncompacted DAW stored as piles of mater_ial equivalent to the containers listed below. Administrative limits will be placed on the activity content of combustibles available for a fire to limit the fire severity and to prevent having a fire that will cause a release corresponding to the dos~. .associated with two times the ODCM limit at a location corresponding to the highest atmospheric dispersion factor listed in the ODCM or above. This will ensure that the release levels will remain below the NOUE limit for the site. Procedures will be developed .in accordance with the Radiation Protection Program to perform monitoring_ . .and to implement the source term administrative limits to ensure potential events do not exceed site boundary dose limits. The following Table_ provides an example of these administrative limits from Attachment 2. Uncontained DAW Elevated Release No HEPA 100 cm Container Type 1.27 cm mR/hr 30 cm mR/hr mR/hr B-25 5.11 E+02 2.79E+02 9.63E+01 20-yd Roll Off 9.88E+01 6.40E+01 3.20E+01 30-yd Roll Off 7.90E+01 5.72E+01 3.08E+01 1 Defueling, MUP, and Normal refer to a specific mix of radionuclides and are not an indicator of the origin of the contamination. 3

Attachment 1 TMl2-RA-COR-2023-0002 I40-yd Roll Off 6.46E+01 5.06E+01 2.93E+01 In addition, this new analysis establishes cut-off dose rate limits for the building installed HEPA Filters such that in the event of a 100% HEPA Filter failure and release of entrained contamination due to a fire, the criteria for an NOUE would not be exceeded. These limits are: 

                               ,Distance from      Installed HEPA External Surface        Filters Dose of HEPA Filter             Rate Array, cm         Limit, mR hr-1 1.27                 104 30                   42 100                   14 Procedures prepared in accordance with the Radiation Protection Program will be developed to perform this monitoring and to impl~ment the ~dn.,_inistrative limits described above.

Additionally, 

                .the analysis in Attachment 2 evaluates the off-site dose consequences from a fire in a contaminated area. The highest contamination levels were found on the 305 ft. elevation at the open hatch to the air cooler. The Cs-137 levels were on the order of several hundred thousand dpm/100 cm 2 . This evaluation clearly shows that surface
  • contamination in the fire zone areas is not likely to be a significant factor relative to the activity limits on combustible waste materials. This analysis shows that for a potential 10,000 ft2 fire in the Reactor Building, the NOUE criteria would not be exceeded.

In addition to the events analyzed in Attachment 2, other events that could exceed the threshold of an NOUE were reviewed based on the PDMS SAR, NUREG/CR-0130 "Technology, Safety, and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station" (Reference 3), NUREG-0586 "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities" (Reference 4), NUREG/CR-2601 "Technology, Safety and Costs of Decommissioning Reference Light Water Reactors Following Postulated Accidents" (Reference 5), and NUREG-0683 "TMl-2 Programmatic Environmental Impact Statement" (Reference 6). The events that could exceed the threshold of an NOUE were determined to be an oxyacetylene explosion, a drop of a spent resin liner, and a failure of a Processed Water Storage Tank (PWST). With respect to an oxyacetylene explosion, there are no current plans to use oxyacetylene in the vicinity of high source term components in the Reactor Building (i.~., the Reactor Vessel Internals and the Reactor Coolant System and associated components). For these high source term components, we plan to use mechanical 4

Attachment 1 TM 12-RA-COR-2023-0002 means for size reduction. Additionally, any use of oxyacetylene in the vicinity of high source term components in the Reactor Building would be evaluated and controlled by the TMl-2 Engineering Program and be subject to a review in accordance with 10 CFR 50.59 to confirm prior NRC approval is not required. With respect to a drop of a spent resin liner, an analysis was provided to the NRC in Reference 7 which analyzed the drop of a Zeolite liner. This scenario assumed that the water processing system would be housed in the Chemical Cleaning Building (CCB) and expended liners would be transferred from the CCB to the Turbine Building and an outside drop and breach of a liner was possible. Since performance of that analysis, a decision has been made to locate the proposed water processing system in the HEPA Filtered Fuel Handling Building. Thus, the earlier concern with an outside unfiltered release has been eliminated. With respect to a failure of a PWST, an analysis was provided to the NRC in Reference 7 which established radionuclide activity limits for these tanks. In addition to the above, Reference 8 requested information on three other potential events: Buildup of Radiolytic Gases, Dust Explosion and Exothermic Reaction Hazard, and a potential fire in the cork seam. These events were addressed in Reference 9. In summary, the analysis results for a Reactor Building fire, using administrative controls specified in plant procedures, demonstrate that radiation exposure will not exceed two times the ODCM limit at the site boundary. The revised dose calculation is provided as Attachment 2 and has been independently reviewed to ensure the calculation has transparent, traceable source terms and a clear discussion of the airborne fraction released. In addition, other potential events have been considered and addressed as indicated above. References

1. Letter TMl2-RA-COR-2022-0021, "License Amendment Request - Three Mile Island, Unit 2, Decommissioning Technical Specifications, Supplement to Response to Request for Additional Information," dated October 31, 2022 (ML22307A082)
2. Calculation 6612-93-021 Rev. 0, "TMl-2 Waste Stream Update," dated June 25, 1993.
3. NUREG/CR-0130, "Technology, Safety, and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station," 1978.
4. NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, Supplement 1" November 2002 (ML023470304; ML023470323)
5. NUREG/CR-2601, "Technology, Safety and Costs of Decommissioning Reference Light Water Reactors Following Postulated Accidents (ML14023A049; ML14023A050).
6. NUREG-0683, "Programmatic Environmental Impact Statement related to decontamination and disposal of radioactive wastes resulting from March 28, 1979 5

Attachment 1 TM 12-RA-CO R-2023-0002 accident at Three Mile Island Nuclear Station, Unit 2" (ML20003C732; Supplement 1, ML20106J132; Supplement 2, ML20235A112; Supplement 3, ML20247F778).

7. Letter TMl2-RA-COR-2022-0007, "License Amendment Request- Three Mile Island, Unit 2, Decommissioning Technical Specifications, Response to Questions,"

dated April 8, 2022.

8. NRC Letter, "Request for Additional Information for Requested Licensing Action Regarding Decommissioning Technical Specifications," dated July 29, 2022 (ML22210A087; RAI Enclosure ML22210A088)
9. Letter TMl2-RA-COR-2022-0019, "License Amendment Request - Three Mile Island, Unit 2, Decommissioning Technical Specifications, Response to Request for Additional Information," dated September 29, 2022 (ML22276A024) 6

Attachment 3 TM 12-RA-COR-2023-0002 ATTACHMENT 3 Proprietary Information Affidavit 1 Page Follows

TMl-2 Solutions Proprietary Information Affidavit, Affidavit of Amy c. Hazeihoff; Vic~ Rtesigent Regulatory Affairs, TMl,,.2 Solution~; LlC. Trlill-2 Solutions; LL.c With respec_t tQ tfie* following document, which is the prop~lfy :Qf TMl-i Soiutionf?, *4~G. *an!=! whi:Ch il:!. provided in :support ofthis Li9ense Amendme.nt R~quest: 164.090.:EN.,CAL~oo4*"source Term Umitations* and Adminlstrative Controls for the TM1-2* DeGQrnr:nis$i.(minQ Emergency Plan ActiQ11 Levels" This :document consists of Rt6prietary information that TMl_;2 Solution~. ~LC cc1.r:isider~ *c.¢rifideotial. 

,Rele~Se of t!:tis informatio.ii wolild cause irreparable harm to the ~Qr!'i~itive position Qf TMI..:2 Solutions, LLC. This ba$is fpi: this declaration is:

I. This information is owned and mainµ:iined a~ prop~ie~ry !:>y TMl-2 Solutions,. LLC, II. This information is routinely 'held il'l.CClllfrd'/1!1Ce by Tl'llll_;2 Solutions, LLC, and. not disclose~:Uo the public,

  • Ill. This inform1;1tidn ,~.being reqi.18$te:o to. be held in confidence by the, NRQ ~Y ttii.s peti~icm, W. This infi;!rm~ti6n is not available in public s:ources, V. This information would .cause:substantial h~.rm tP Tl\lll ..2 S.6IUtions, LLC, if'it were

,released publicly,. a11d VI: the information to be :Witnh~ld vva$ transmitted to the: NRC in confidence. I, Amy C. H1;1zelhof( b~il'19 d4lysW6tn, state that I am the personwho subscribes my nart1e to the foregoing statememt, I am authorized to execute the Affii:laviton behc1_lf of trv'!l.-2 Solµtioris, LLC, ai:ld lhat the rtiatters a11d facf1:1,set forth in the statement are true to (he best of n,y ~h0Wledge1 ihfofmatioh, and. belief: * * * *

  • A.my c. l-f~ei.'1off Vice Pr~sic;tent Regulatory Affairs TM1.:2 SQll.ltiohs, LLC ,
                                                          -th Silvorri To And Su.bs,cribed Before Me This;             t Day of        0a f.1Uf.i. try   'J.,C(~ 3 My Commission Expires.           tJ eeA:?.i,;,,,b-.eir JJ,.; iooi 7 w~,.,,,,{I ..-v.-p.~                   '71~-;A-'-{"1~

Attachment 4 TMl2-RA-COR-2023-0002 ATTACHMENT 4 Revision to RAI 10-13 Responses 11 Pages Follow

Attachment 4 TM 12-RA-COR-2023-0002 MATERIAL CONTROL AND ACCOUNTING (MC&A) RAI 10 Accounting for Debris Material Comment: It is not clear in the LAR how TMl-2 Solutions plans to control and account for Debris Material throughout decommissioning. Basis: Debris Material must be controlled and accounted for at all times during decommissioning because the Debris Material contains large quantities of SNM, including uranium-235 and plutonium. Once TMl-2 has entered DECON, the applicant has stated that SNM will be retrieved, aggregated, and placed into dry cask storage using various shapes and sizes of containers to place into a basket and canister. To minimize aggregating the remaining SNM, the core debris will be generally packaged and loaded as it is retrieved. These canisters will then be transferred to the expanded Independent Spent Fuel Storage Installation (ISFSI) inside the Three Mile Island Station, Unit No. 1 ("TMl-1 "), ISFSI fence to store the canisters after TMl-1 completes their spent fuel transfer campaign to the ISFSI. In addition, estimates of the quantities and form of SNM at TMl-2 provided by the applicant indicate that the site may need more detailed plans for material control and accounting during decommissioning, compared to sites where SNM is generally restricted to undamaged spent fuel assemblies. 10 CFR Part 74, "Material Control and Accounting of Special Nuclear Material," establishes requirements for the control and accounting of SNM at fixed sites and for documenting the transfer of SNM. General reporting requirements as well as specific requirements for certain licensees possessing SNM of low strategic significance, special nuclear material of moderate strategic significance, and formula quantities of strategic special nuclear material are included. Path Forward: Describe how TMl-2 Solutions will control and account for Debris Material being removed from the Reactor Building to the Three Mile Island ISFSI throughout the decommissioning process in order to meet the applicable requirements of 10 CFR Part 74. Describe TMl-2 plans to refine current rough estimates of radionuclide content in Debris Material in existing reports and provide more accurate information on quantities of SNM as materials are packaged and removed. TMl-2 Solutions RAI 10 Response: Material Control and Accountability Program

Description:

TMl-2 Solutions has developed a program to ensure proper accounting of SNM is conducted throughout the decommissioning process, meeting the applicable requirements of 10 CFR 74, with applicable exemptions. Current rough estimates of radionuclide content will be refined using Low-Level Radioactive Waste (LLRW) characterizations and the site final status survey as explained below. 1

Attachment 4 TM 12-RA-COR-2023-0002 Overview of TMl-2 NMCA Processes SNM is a constituent of Fuel Bearing Material and is dispersed as shown in the discrete quantities of U02 identified in Tables 4.3-1 and 4.3-2 of the PDMS SAR (Reference 1). Each of these quantities of fuel was estimated during post defueling survey reports and, together with nine discrete and packaged items turned over from TMl-1, represent the material balance and starting inventory for the TMl-2 Decommissioning Project reported annually (Reference 2). The accident at TMl-2 caused a significant portion of the fuel to derange, melt, oxidize, and combine with core materials and the total amount of post-accident material was estimated to be 133,000 Kgs (~293,000 pounds) (Reference 3). TMl-2 was exempted from multiple SNM Control and Accountability requirements associated with transfer and disposal, including physical inventory (Reference 4). In lieu of the requirements, TMl-2 provided information describing the physical contents of the shipments made to DOE. Following completion of damaged fuel removal, post-defueling characterization surveys identified 1097 kg of U02 remaining in various locations in TMl-2 with a total uncertainty of+/- 40% (References 5-12). The capability to significantly reduce the 40% uncertainty would require characterizing the collected fuel debris in each container using sophisticated hot cell and laboratory facilities with the means to homogenize, sample, weigh, and analyze the contents of each canister. Such facilities did not (and do not) exist at TMl-2. The results of the post defueling survey reports were reviewed and approved by the NRC in Reference 13. 2

Attachment 4 TM 12-RA-COR-2023-0002 TMl-2 NMCA Process Figure 1 provides a high-level flow chart for MC&A requirements from the existing configuration through the decommissioning project. Figure 1 SNM Identification SNM Movement from In Situ to Package Interim FBM Canister LLRW Sample Storage 

                     .i              .i              i            .i Preliminary SNM
                                            ~     Final SNM Accounting Accounting
                            .i                             t Annual Inventory Annual Reporting SNM Identification:
  • Review each discrete location containing fuel bearing material to determine optimal decommissioning sequences using engineering assessments created per written procedures. The engineering assessments integrate inputs of contemporaneous regulatory waste classification and shipping requirements with known historical reports and data. The engineering assessments generate operational assumptions and constraints, cut and packaging plans, expected waste packages, waste classification, estimated SNM in waste packages and Dry Storage Canisters, evaluate 10 CFR 37 radioisotopes of concern in resulting packages, evaluate whether confirmatory radiological measurements are required, and describe the approach for characterization of low-level radioactive waste (LLRW) packages.

SNM Movement from In Situ to a Package:

  • The starting in situ physical condition of SNM as a constituent of Fuel Bearing 3

Attachment 4, TMl2-RA-COR-2023-0002 Material (FBM) is that SNM has been best-estimated using state of the art techniques, many of which were developed specifically for conditions at TMl-2 in post defueling survey reports (References 5-12). Uncertainties associated with individual discrete estimates vary between 17-104% and weighted average to a total of+/- 40% uncertainty. Physical inventories have not occurred since the post defueling surveys were completed because they were exempted (Reference 4).

  • In-situ physical configuration of FBM varies and will generally be one of: films or fines plated onto previous water to metal interfaces, loose debris or gravel-like particulate of varying size, artifacts which are tightly adhered to larger plant components, or captured in varying types of filter media.
  • Physical inventory of SNM as a constituent of FBM in situ is not required under the current exemption (Reference 4 ).
  • SNM movement from its in-situ condition to a confined package shall be to one of:

o Interim Storage: FBM which is planned to be disposed in a Dry Storage Canister and is removed from its in-situ configuration prior to availability of a Dry Storage Canister shall be placed in an approved interim storage location per written procedure. o FBM Canister: FBM which will be removed from its in-situ configuration and directly placed into Dry Storage Canisters per written procedures. o LLRW: It is anticipated much of the waste produced at TMl-2 will contain trace amounts of SNM. LLRW packages which have been characterized for waste class, shipping criteria, and disposal site waste acceptance criteria and do not contain 1 gram or more of SNM do not require control and accountability actions. LLRW waste packaging procedures which contain 1 gram or more of SNM will require material accountability and control actions per written procedures. o Sample: It is anticipated most samples produced at TMl-2 and sent off site for analysis will contain trace amounts of SNM. While samples generally expected to contain very small total quantities of materials, any sample which contains 1 gram or more of SNM will require material control and accountability actions. Samples which do not contain 1 gram or more of SNM will not require control and accountability actions. Preliminary SNM Accounting:

  • SNM as a constituent of FBM which has been either placed in interim storage

_containers or into Dry Storage Canisters will have preliminary SNM accounting performed per a written procedure to analytically determine SNM content based on the engineering assessment process. This will remain preliminary until after the final status survey for the decommissioning project is complete. 4

Attachment 4 TMl2-RA-COR-2023-0002 Final SNM Accounting: Accountability of SNM as a constituent of FBM disposed in Dry Storage Canisters will be based on the net of starting inventory reported in the PDMS SAR (Reference 1) less SNM disposed in LLRW or sent as a Sample and reconciled using the TMl-2 site Final Status Survey data. This is similar to the reporting method following the defueling effort in 1990.

  • Final Material Transaction Reports will be performed per a written procedure for LLRW and Sample shipments containing 1 gram or more of SNM.
  • The Final Status Survey will validate via an engineered sampling scheme that any material remaining on site is below the derived concentration guideline levels (DCGLs).
  • The final SNM accounting will reconcile the total SNM shipped and the SNM loaded into dry storage canisters to the final status survey. Final material balance reports will be generated at that time by written procedure.

Annual Inventory: An inventory will be conducted per a written procedure annually as follows:

  • SNM as a constituent of FBM which has not been removed from its in-situ configuration during the decommissioning process will not be physically inventoried. Remaining in-situ material will be tracked per written procedure through decommissioning processes.
  • SNM in any packages physically on site (to include LLRW or samples awaiting shipment, interim storage of FBM, and FBM in Dry Storage Canisters) at the time of the annual inventory will be physically inventoried.

Annual Reporting: TMl-2 will report the SNM inventory on-site annually using the results of the annual physical inventory (as determined above), analytically evaluated remaining in-situ material and shipments (LLRW and samples) containing SNM per written procedure. RAI 10 Specific Response The Program Description above describes how TMl-2 Solutions will control and account for Debris Material being removed during decommissioning processes in a manner compliant with 10 CFR 74 with existing exemptions. For material being removed from TMl-2 systems which will be moved to dry cask storage, each canister will be analytically evaluated per written procedures based on the estimates of record (References 5-12) and an estimated quantity of SNM will be assigned which is associated with estimates from the cargo being packaged. These estimates for each dry canister will not improve on existing uncertainties. The remaining TMl-2 fuel bearing material represents approximately two standard spent 5

Attachment 4 TM 12-RA-COR-2023-0002 fuel modules of fuel and fission product materials and will be stored within approximately 14 dry storage canisters; each dry storage canister is designed to contain up to 37 spent fuel modules. TMl-2 fuel bearing material will be aggregated to quantities significantly lower than the dry storage systems at most reactor plants. The SNM estimate for each dry canister will be considered as a preliminary estimate and will be refined by, and finalized, upon completion of the TMl-2 decommissioning against the small quantities shipped as samples, LLRW and the Final Status Survey. For low quantities of fuel bearing material being removed from TMl-2 systems which will be disposed as sample materials or within low level radioactive waste, the waste characterization process required by 10 CFR 61 will be per written procedure and will improve upon the current estimates with additional sampling and characterization processes. This will not have an appreciable effect on the refinement of the estimates for the material in the DCSs. References

1. Three Mile Island Nuclear Station, Unit 2 (TMl-2), "Revised Update 14 of Post-Defueling Monitored Storage Safety Analysis Report," TMl2-RA-COR-2021-0014, dated September 29, 2021
2. Exelon Generation, "Transmittal of DOE/NRC Form 742, Material Balance Report for Three Mile Island," NF210273, dated August 16, 2021
3. Three Mile Island Nuclear Station, Unit 2 (TMl-2), Defueling Completion Report, Final Submittal," 4410-90-L-0012, February 22, 1990 (ML20011F539)
4. US NRC "Approval of Exemption from 10 CFR 30.51, 40.61, 70.51 (d), and 70.53" dated October 17, 1985 (ML20138D392)
5. Three Mile Island Nuclear Station, Unit 2 (TMl-2), "SNM Accountability," 4410 L-0162, dated September 30, 1988 (ML20207M226)- (Reactor Vessel Plenum)
6. Three Mile Island Nuclear Station, Unit 2 (TMl-2), "SNM Accountability," 4410 L-0097, dated September 22, 1989 (ML20248B795)- (Letdown Coolers, Pressurizer, Reactor Building Basement)
7. Three Mile Island Nuclear Station, Unit 2 (TMl-2), "SNM Accountability," 4410 L-0019, dated March 14, 1990 (20012D159)-(ReactorVessel Head)
8. Three Mile Island Nuclear Station, Unit 2 (TMl-2), "SNM Accountability," C312-91-2045, dated June 6, 1991 (ML20077F496) - (Auxiliary and Fuel Handling Buildings)
9. Three Mile Island Nuclear Station, Unit 2 (TMl-2), "SNM Accountability," C312-91-2052, dated June 18, 1991 - (Reactor Building Miscellaneous Components)
10. Three Mile Island Nuclear Station, Unit 2 (TMl-2), "SNM Accountability," C312-91-2055, dated July 3, 1991. (ML20082A241) - (Reactor Coolant System) 6

Attachment 4 TMl2-RA-COR-2023-0002

11. Three Mile Island Nuclear Station, Unit 2 (TMl-2), "SNM Accountability," C312-91-2064, dated August 20, 1991 - ('A' and 'B' Once-Through Steam Generators (OTSGs))
12. Three Mile Island Nuclear Station, Unit 2 (TMl-2), "SNM Accountability," C312 -

93-2004, dated February 1, 1993 - (Reactor Vessel)

13. US NRC, "Post-Defueling Survey Report Reviews" dated November 4, 1994 (ML20078H309) 7

Attachment 4 TM 12-RA-COR-2023-0002 RAI 11 Reports of Loss or Theft of SNM Comment: In the LAR TMl-2 Solutions does not address reporting of loss, theft, or attempted theft of SNM. Basis: 10 CFR 74.11 (a), "Reports of loss or theft or attempted theft or unauthorized production of special nuclear material," requires each licensee who possesses one gram or more of contained uranium-235, uranium-233 or plutonium to notify the NRC Operations Center within 1 hour of discovery of any loss or theft or other unlawful diversion of SNM which the licensee is licensed to possess, or any incident in which an attempt has been made to commit a theft or unlawful diversion of SNM. Path Forward: Provide a description of the MC&A activities that are performed or the measures in place to show how the reporting requirement of 10 CFR 74.11(a) is met. TMl-2 Solutions RAI 11 Response: The TMl-2 Materials Security Plan establishes a security zone for which:

  • Personnel access is controlled
  • Random searches of personnel and equipment upon exiting the security zone are conducted. These random searches include radiation detection which will detect radioisotopes present with SNM.

Security personnel are notified upon detection of loss, theft, or unlawful diversion of SNM and upon any incident in which an attempt has been made to commit a theft or unlawful diversion of SNM. The TMl-2 Materials Security Plan is implemented using security personnel under contract from TMl-1. Security notifications of less than 4 hours are made by the TMl-1 security organization using the TMl-1 processes and procedures. TMl-2 has the requirements of 10 CFR 74.11 (a) included in TMl2-RA-PR-005, Reporting of Events and Conditions. This procedure details the one-hour reporting requirement and meets 10 CFR 74.11(a). 8

Attachment 4 TMl2-RA-COR-2023-0002 RAI 12 Material Status Reports Comment: In the LAR TMl-2 Solutions does not address completion or submission of Material Balance Reports or Physical Inventory Listing Reports. Basis: 10 CFR 74.13(a), "Material status reports," requires each licensee possessing SNM in a quantity totaling 1 gram or more of contained uranium-235, uranium-233, or plutonium to complete and submit, in computer-readable format Material Balance Reports concerning SNM that the licensee has received, produced, possessed, transferred, consumed, disposed, or lost. The Physical Inventory Listing Report must be submitted with each Material Balance Report. Path Forward: Provide a description of the MC&A activities that are performed or the measures in place to show how the reporting requirements of 10 CFR 7 4.13(a) are met. TMl-2 Solutions RAI 12 Response: The Reference 1 exemption required the preparation of Material Balance Reports and Physical Inventory Listings of remaining SNM at TMl-2. Reference 2 provided a report by the NRC reviewing and accepting the post-defueling survey report results, including the overall uncertainty of+/- 40%. Accounting for that uncertainty, TMl-2 Solutions will execute reporting to written procedures in two stages: Preliminarily for material which is packaged for dry storage at an onsite ISFSI during decommissioning processes, and finally, for both fuel bearing material packaged for samples or disposal as low-level radioactive waste and for packaged dry storage containers upon completion of decommissioning activities reconciled using final status survey data. Preliminary SNM Accounting: SNM as a constituent of FBM which has been either placed in interim storage containers or into Dry Storage Canisters will have preliminary SNM accounting performed per a written procedure to analytically determine SNM content based on the engineering assessment process. This will remain preliminary until after the final status survey for the decommissioning project is complete. Final SNM Accounting: Accountability of SNM as a constituent of FBM disposed in Dry Storage Canisters will be based on the net of starting inventory reported in the PDMS SAR (Reference 3) less SNM disposed in LLRW or sent as a Sample and reconciled using the TMl-2 site Final Status Survey data. This is similar to the reporting method following the defueling effort. Final Material Transaction Reports will be performed per a written procedure for LLRW and Sample shipments containing 1 gram or more of SNM. The Final Status Survey will validate via an engineered sampling scheme that any material remaining on site is below the derived concentration guideline levels (DCGLs). The final SNM accounting will reconcile the total SNM shipped and the SNM loaded into dry storage canisters to the final status survey. Final material balance reports will be generated at that time by written procedure. 9

Attachment 4 TM 12-RA-COR-2023-0002 References

1. US NRC "Approval of Exemption from 10 CFR 30.51, 40.61, 70.51 (d), and 70.53" dated October 17, 1985 (ML20138D392)
2. US NRC, "Post-Defueling Survey Report Reviews" dated November 4, 1994 (ML20078H309)
3. Three Mile Island Nuclear Station, Unit 2 (TMl-2) "Revised Update 14 of Post-Defueling Monitored Storage Safety Analysis Report," TMl2-RA-COR-2021-0014, dated 29 September 2021 10

Attachment 4 TMl2-RA-COR-2023-0002 RAI 13 Nuclear Material Transaction Reports Comment: In the LAR TMl-2 Solutions does not address completion of Nuclear Material Transaction Reports. Basis: 10 CFR 74.15(a), "Nuclear material transaction reports," requires each licensee who transfers or receives SNM in a quantity of 1 gram or more of contained uranium-235, uranium-233, or plutonium to complete, in computer-readable format, a Nuclear Material Transaction Report. In addition, each licensee who adjusts the inventory in any manner, other than for transfers and receipts, shall submit a Nuclear Material Transaction Report, in computer-readable format, to coincide with the submission of the Material Balance Report. Each licensee who transfers SNM shall submit a Nuclear Material Transaction Report no later than the close of business the next working day. Each licensee who receives SNM shall submit a Nuclear Material Transaction Report within 10 days after the material is received. Path Forward: Provide a description of the MC&A activities that are performed or the measures in place to show how the reporting requirements of 10 CFR 74.15(a) are met. TMl-2 Solutions RAI 13 Response: As discussed in RAI 10, Material Transaction Reports for all packages containing 1 gram or more of SNM which are shipped from TMl-2 will be created in accordance with written procedures. For LLRW waste packages, characterization prior to the shipment will meet 10 CFR 74.15(a) requirements. For SNM in dry cask storage, characterization will meet 10 CFR 74.15(a) requirements after completion of final status survey for TMl-2 per written procedure. 11

Attachment 5 TM 12-RA-COR-2023-0002 ATTACHMENT 5 Evaluation of the proposed changes to the TMl-2 POL and TS and Revised NSHC Attachment 5 provides a revised evaluation of certain proposed changes to the TMl-2 POL and TS that have been affected by References 2-10, and the revised No Significant Hazards Consideration (NSHC). 37 Pages Follow

Attachment 5 TMI2-RA-COR-2023-002 Licensee Identified Supplemental information to the LAR Reference 1 requested an amendment to the Possession Only License (POL) and Appendix A, Technical Specifications (TS), of POL No. DPR-73 ("License") for Three Mile Island Nuclear Station, Unit 2 ("TMl-2"). This proposed License Amendment Request (LAR), upon approval, will revise the POL and the associated TS to support the transition of TMl-2 from a Post-Defueling Monitored Storage (PDMS) (equivalent to SAFSTOR) condition to that of a facility undergoing radiological decommissioning (DEGON) pursuant to 10 CFR 50.82(a)(7). Since the submittal of Reference 1 certain of the proposed Technical Specification changes have been modified as described below TS Section 1.0 Definitions Definitions described in TS Section 1.0, "Definitions," are either retained in the Technical Specifications, proposed for deletion since they are only relevant to TMl-2 during the PDMS condition or proposed for relocation to the Decommissioning Quality Assurance Program (DQAP). This change is administrative in nature and does not impact nuclear safety. The standard convention of indicating the defined term in ALL CAPITAL LETTERS throughout the TS has been adopted in the DTS. 1 of37

Attachment 5 TMI2-RA-COR-2023-002 Definitions Basis for Deletion POST-DEFUELING MONITORED This definition is proposed for deletion STORAGE (PDMS) since the term is not used in any Technical Specificati.on. Definition Deleted: 1.2 POST-DEFUELING MONITORED STORAGE (PDMS) is that condition where TMl-2 defueling has been completed, the core debris removed from the reactor during the cleanup period has been shipped off-site and the facility has been placed in a stable, safe, and secure condition. Definitions Basis for Retention/Relocation OFF-SITE DOSE CALCULATION Basis for Retention MANUAL This definition is cited in Technical Definition Retained: Specification 6.8.1.2, "Annual Radioactive 1.12 OFF-SITE DOSE CALCULATION Effluent Release Report." As suggested in MANUAL (ODCM) shall contain the Reference 3 RAl-8, the Reference 6 methodology and parameters used in the response to RAl-8 committed to calculation of off-site doses resulting from supplement the LAR to retain Technical radioactive gaseous and liquid effluents, in Specification 6.8.1.2. Thus, as Technical the calculation of gaseous and liquid Specification 6.8.1.2 "Annual Radioactive effluent monitoring alarm/trip setpoints, Effluent Release Report." Is being and in the conduct of the Radiological retained this definition which supports Environmental Monitoring Program. The Technical Specification 6.8.1.2 will also be ODCM shall also contain descriptions of retained. the information that should be included in the Annual Radioactive Effluent Release Reports required by Specifications 6.8.1.2. 2 of37

Attachment 5 TMI2-RA-COR-2023-002 Definition Relocated: Basis for Relocation 1.12 OFF-SITE DOSE CALCULATION This definition is cited in Technical MANUAL (ODCM) shall contain the Specification 1.18, "Site Boundary"; methodology and parameters used in the Technical Specification Section 6.7.4.a, calculation of off-site doses resulting from "Radioactive Effluent Controls Program"; radioactive gaseous and liquid effluents, in Technical Specification Section 6.7.4.b, the calculation of gaseous and liquid "Radiological Environmental Monitoring effluent monitoring alarm/trip setpoints, Program"; Technical Specification 6.8.1.1, and in the conduct of the Radiological "Annual Radiological Environmental Environmental Monitoring Program. The Operating Report";; Technical ODCM shall also contain (1) the programs Specification 6.9, "Records Retention"; required by the Decommissioning Quality and Technical Specification Section 6.12, Assurance Plan and (2) descriptions of the "Offsite Dose Calculation Manual information that should be included in the (ODCM)"; which, are proposed for Annual Radiological Environmental relocation to the DQAP. Operating and Annual Radioactive Therefore, the definition of ODCM is Effluent Release Reports required by the proposed for relocation to the DQAP for Decommissioning Quality Assurance Plan these Sections. and Specification 6.8.1.2, respectively. 3 of37

Attachment 5 TMI2-RA-COR-2023-002 Definitions Basis for Relocation NPDES PERMIT The proposed change is to delete this 1.19 The NPDES PERMIT is the National definition from the current TMl-2 TS and Pollutant Discharge Elimination System relocate it to the DQAP. The NPDES (NPDES) Permit No. PA0009920, PERMIT is cited in TS 6.13.2, "Exceeding effective January 30, 1975, Issued by the Limits of Relevant Permits." TS 6.13.2 is Environmental Protection Agency to proposed for relocation to the DQAP. Metropolitan Edison Company. This Therefore, the definition of NPDES permit authorized Metropolitan Edison PERMIT is proposed for relocation to the Company to discharge controlled DQAP. wastewater from TMI Nuclear Station Into The NPDES permit language is updated the waters of the Commonwealth of from the priginal submittal in Reference 1 Pennsylvania. to delete Exelon Generation Company, LLC and to reflect current assignment of the permit to Constellation Energy Generation LLC. The NPDES PERMIT is the National Pollutant Discharge Elimination System (NPDES) Permit No. PA0009920, effective June 1, 2010January 30, 1975, issued by the Pennsylvania Department of Environmental Protection Agenoy and assigned to Exelon Generat5on CompaR)<, LLC. Constellation Energy Generation, LLC . Metropolitan Edison Company. This permit authorizes discharge to the Susquehanna River with effluent limitations, monitoring requirements and other conditions within the permit. Metropolitan Edison Company to disoharge oontrolled 1Nastewater from TMI Nuolear Station Into the 1Naters ef the Commonwealth ef Pennsylvania. 4 of37

Attachment 5 TMI2-RA-COR-2023-002 TS SECTION 3/4.1 CONTAINMENT SYSTEMS TS Section 3/4.1 contains Limiting Conditions for PDMS that assures that the containment is maintained as a contamination barrier for the residual contamination which remains inside the containment. This section is proposed for deletion in its entirety. Control of residual contamination inside of containment will be provided by the Radiation Protection Program (RPP) and through implementation of procedures which address execution of Decontamination and Decommissioning (D&D) work and support activities, including personnel safety as well as measures to maintain occupational dose As Low As Reasonably Achievable (ALARA) and below the occupational dose limits in 10 CFR Part 20 during decommissioning as applicable. Procedures associated with Phase 1b will be developed to retrieve the remaining core debris and decontaminate high radiation areas. Phase 2 procedures will also be developed; however, the focus of these procedures is related to performing D&D operations in a facility which has not experienced an accident. The RPP and associated implementing procedures will control the residual contamination located inside of the containment and therefore achieves the same results as the containment contamination barrier during the PDMS condition. Therefore, the containment isolation Technical Specification Limiting Conditions for PDMS are not required, and do not apply in Phase 1b and Phase 2. Since the submittal of Reference 1 the basis for deletion of this Technical Specification has changed based on the event analyses provided in response to NRG Requests for Additional information in References 2 and 3 Current TMl-2 TS Basis for Deletion 3.1.1.1 Containment This Technical Specification is proposed for deletion. Isolation During PDMS, containment isolation is maintained to assure the containment is properly maintained as a contamination barrier for the residual contamination which resides inside the containment. 5 of37

Attachment 5 TMI2-RA-COR-2023-002 Current TMl-2 TS Basis for Deletion 3.1.1.1 Containment Ne majeF EleeemmissieRiR§ will eee1::1F El1::1FiR§ Pl=lase 1a. +l=le Isolation Pl=lase 1a eeRElitieR is a eeRtiR1::1atieR et: tl=le PQMg eeRElitieR. Wl=lile iR Pl=lase 1a, +Ml 2 g9h,1tieRs eem13lies witl=I tl=le PQMg ::i:g as FeviseEI ey tl=le iss1::1aRee et: tl=le LieeRse +rnRsfoF AmeRElmeRt wl=liel=I was a1313mveEI ey tl=le NRG iR Ref:ernRee 23. J;:eF tl=le PQMg eeRElitieR, GP61 N1::1eleaF 13eFfuFmeEI aR 1::1RaRtiei13ateEl eveRts aRalysis as weseRteEI iR A1313eREli* l=I, geetieR g_2 et: tl=le PQMg gaf:ety ARalysis Re130Ft (gAR). 

                    +l=le 131::1F13ese et: tl=le aRalysis waste EleteFmiRe tl=le 1::1RaRtiei13ateEI eveRt tl=lat 13mEl1::1ees tl=le 001:mEliR§ FaEliele§ieal Elese at tl=le site ee1::1RElai=y El1::1FiR§ PQMg_ +l=le Fes1::1lts et: tl=lis aRalysis iRElieate tl=lat a f:iFe iR tl=le RB i.vitl=I tl=le RB veRtilatieR aREI 131::1F§e system iR e13eFatieR is tl=le 1::1RaRtiei13ateEI eveRt tl=lat weE11::1ees tl=le ee1::1REliR§ rnEliele§ieal Elese at tl=le site ee1::1RElal)' El1::1FiR§ PQMg_
                    +l=le f:iFe iRsiEle et: tl=le RB witl=I tl=le RB veRtilatieR aREl 131::1F§e iR e13eFatieR was eval1c1ateEI ey tl=le NRG as 13aFt et: tl=le e*eleR Fe~1c1est feF e*em13tieR f:Fem 13eFtieRs et: 1Q GJ;:R aQ.47 aREI 1Q GJ;:R aG, A1313eREli* e (Ref:eFeRee g)_ +l=le Fes1c1lts et: tl=le ~JRG eval1::1atieR eeHfiFm tl=lat tl=le +Ml 2 f:aeility we1c1IEI Ret l=lave eeRse~1c1eRees tl=lat ee1c1IEl 13eteRtially e*eeeEI tl=le a1313lieaele Elese limits iR 1Q GJ;:R 1QQ.11 aREI tl=le ePA PAGs (RefoFeRee 11 ).

AfteF tl=le iss1c1aRee et: Ref:eFeRee 11, tl=le FiFe PFeteetieR Pm§Fam eval1c1atieR,Re 1.iisieR 2, (Ref:eFeRee 9) 1Nl=liel=I seFVeEl as aR iR131c1t te e*eleR's e*em13tieR Fe~1c1est (Ref:eFeRee g) was FeviseEI aREI Feiss1::1eEl as FevisieR 13 (Ref:eFeRee 12). RevisieR 13 et: tl=le Fi Fe PmteetieR Pm§Fam e 1.ial1c1atieR 1c1ses 1c113ElateEI se1::1rne teFm iRfeFmatieR 1Nl=liel=I asse1::1Rts feF 2@ yeaFs et: Eleeay (1992 tl=IFel::l§R 2Q1 g) as 1Nell as aese1c1RtiR§ feF aEIElitieRal leese eeRtamiRatieR. J;:eEleFal G1c1iElaRee Re130Fts 11 aREl 12 aFe a1313lieEl feF Elese 69RVeFsieR f:aetern. +l=le rns1::1lts 13rnseRteEl iR RefornRse 12 iRElieate tl=lat tl=le f:iFe iRsiEle et: tl=le RB witl=I tl=le RB veRtilatieR aREI J3l:IF§e iR e13eFatieR FemaiRs tl=le mest limitiR§ 1::1RaRtiei13ateEI eveRt. Qese at tl=le site ee1::1RElai=y Elees Ret e*eeeEI tl=le limits weseRteEI iR 1Q GJ;:R 1QQ.11 aREI tl=le ePA PAGs. Following Phase 1a, TMl-2 will enter Phase 1b and Phase

2. During Phase 1b and Phase 2, major decommissioning activities as defined .in 10 CFR 50.2 will be performed. The TMl2 RPP will address control of residual contamination inside of containment during Phase 1b and Phase 2.

6 of37

Attachment 5 TMI2-RA-COR-2023-002 Current TMl-2 TS Basis for Deletion 3.1.1.1 Containment Development of implementing procedures will take into Isolation account the execution of D&D work and support activities, personnel safety, as well as measures to maintain occupational dose As Low As Reasonably Achievable (ALARA) and below the occupational dose limits in 10 CFR Part 20 during decommissioning as applicable. Procedures associated with Phase 2 will be developed, however the focus of these procedures is related to performing D&D operations in a facility which has not experienced an accident. As described in Reference 4 the basic changes in the Reactor Building (RB) in going from PDMS to D&D is removal of the Equipment Hatch, squaring off of the hole left from Equipment Hatch Removal, and installation of a barrier at the interface between the RB and the outside structure. The RB is a passive radiological barrier during Phase 1b to the extent that the door between the RB and the outside structure would only be open for the period of time necessary to allow passage of material or personnel between the two structures. Additionally, during Phase 1b other openings may be made in the containment structure. Following good ALARA practices these openings will also include passive radiological barriers. The RPP and associated implementing procedures will provide the required level of control for residual contamination at RB openings such as the Equipment Hatch Area and the containment airlocks when in use. IA Sl:lppei=t ef Q&Q, eAgiAeeFeEI assess epeAiAgs iA tR_e RB easemeAt aAEI at tRe RB eeil:lipmeAt RatsR will ee seAstFl:lsteEI. +Re RPP aAEI assesiateEI implemeAtiAg weseEll:IFes 11.iill pFeviEle tRe Feeil:liFeEI level ef seAtFel Aesessai:y te maiAtaiA FesiEll:lal seAtamiAatieA iAsiEle ef seAtaiAmeAt at tRese as well as etReF epeAiAgs Sl:ISR as seAtaiAmeAt aiFlesks. Airborne radiation monitoring will be provided at the containment openings. Limiting Condition for PDMS 3.1.1.1 and associated Surveillance Requirement 4.1.1.1 are proposed for deletion upon entry into Phase 1b and Phase 2. During Phase 1b, the residual contamination inside containment will be controlled by the RPP and adherence to implementing procedures. Features that may be 7 of37

Attachment 5 TMI2-RA-COR-2023-002 Current TMl-2 TS Basis for Deletion 3.1.1.1 Containment employed to manage residual contamination during Phase Isolation 1 b include,

  • Controlling contamination at the source to minimize release and spread inside of containment,
  • Filtration of water and rinsing of items removed from underwater operations.
  • Outside of water, typical contamination control processes include utilization of fixatives on components being dismantled, utilization of water misting during dismantlement to preclude the generation of airborne radioactivity, and decontamination of components prior to dismantlement.
  • Controls that will be implemented for contamination that has been liberated include use of HEPA ventilation, radiological surveys and decontamination of accessible surfaces within containment.
  • Radiological controls will be utilized to maintain worker exposure ALARA and to prevent cross-contamination.
  • Airborne radiation monitoring at the access openings used for detection of residual contamination.
                     ,A.s Eliss1:1sseEl iR SestieR2, aR aRalysis Ras eeeR peFfeFmeEl ef a HIG fiFe ess1:1FFiR§ 01:1tsiEle ef seRtaiRmeRt, 'ltl=lisl=I FepFeseRts tRe mest limitiR§ assiaeRt te 0ss1:1F iR PRase ~ e aREl Pl=lase 2 Felative te Elese at tl=le site e01:1RElaPJ. +l=le Fes1:1lts ef tRe aRalysis i.<eFify tRe plaRt eAB aRa bPi meet tRe ~Q Gi;:R ~QQ.n Elese sFiteFia aREl tRe ePA PAG sFiteFia; eelei.,.< 1NRisR eff site emeFjeRsy FespeRse sapaeilities aFe Ret Feq1:1imEl. +Re aRalysis furtl=leF ElemeRstrates tRat tRe raEliele§isal s0Rseq1:1eRses ef tRe HIG fiFe e01:1REl tRe aese 1

at tRe site e01:1RElaPJ El1:1e te fiFe iRsiae seRtaiRmeRt witR tRe RB veRtilatieR aREl p1:1Fje system iR eperatieR. +Re aese aRalysis assesiatea witR tRe HIG fira Elees Ret take omElit fuF mitigating SSC's. 

                     +ReFe aFe Re pest1:1latea aseiaeRts tRat saR 0ss1:1F iRsiae ef tRe RB a1:1FiR§ PRase ~ e eF PRase 2 tRat Fes1:1lt iR tRe aese at tRe site e01:1RaaPJ e*seeEliR§ tRe limits ef ~ Q GFR ~ QQA ~

aRa tRe ePA PAGs iRsl1:1EliR§ s1:1sR times as wReR tRe seRtaiRmeRt eR§iReeraa assess eq1:1ipmeRt RatsR is epen. 

                     +Re Q&Q pFesess iRsl1:1aes maRy evel1:1ti0Rs tRat will Feq1:1iFe tRe eq1:1ipmeRt RatsR aREl etReF RB assess peiRts te ee 9fMffi 8 of37

Attachment 5 TMI2-RA-COR-2023-002 Current TMl-2 TS Basis for Deletion 3.1.1.1 Containment te alle11o1 mevemeRt ef eEJ1:1i13meRt, i..vaste, aRel etl=leF mateFials Isolation iRte aREI el:lt ef tl=le RB. +l=le RPP will iEleRtify tl=le seRtFels tl=lat i.A1ill ee im13lemeRteEI tl=lFel:l§R 13FeseEll:IFes ell:IFiR§ Q&g astivities essl:IFFiR§ iRsiele ef tl=le RB. Analyses were performed as described in the response to RAl-3 from Reference 3 provided in Attachment 1. These analyses established activity and dose rate limits to maintain the consequences of an off-site release below an Emergency Action Level (EAL) as defined by NEI 99-01 [Reference 7] for permanently defueled reactors (i.e., Notification of an Unusual Event (NOUE) and Alert). Guidance for implementation of administrative controls to meet the limits was also provided. Additionally, the D&D process includes many evolutions that require the equipment hatch and other RB access points to be open to allow movement of equipment, waste, and other materials into and out of the RB. The RPP and the Fire Protection Program Evaluation (FPPE) will identify the controls that will be implemented through procedures during D&D activities occurring inside of the RB. Included is that the door between the RB and the outside structure would only be open for the period of time necessary to allow passage of material or personnel between the two structures. Implementation of these procedures takes into account detailed work planning, and execution of the D&D work and support activities, as well as measures to maintain occupational dose As Low As Reasonably Achievable (ALARA) and below the occupational dose limits in 10 CFR Part 20 during decommissioning. The RPP and associated implementing procedures will control the residual contamination located inside of containment. Therefore, maintaining containment isolation as specified in TS 3.1 .1.1 and the associated surveillance requirement are no longer required and are proposed for deletion. 9 of37

Attachment 5 TMI2-RA-COR-2023-002 Current TMl-2 TS Basis for Deletion 3.1.1.2 Unfiltered Leak This Technical Specification is proposed for deletion. Rate Testing This Technical Specification assures that the unfiltered leak rate from the containment with the RB Breather closed is less than 1/100 of the rate through the RB Breather. 6h:1FiR§ P9MS, aiFBSFRe astivity iR tl=le RB as a Fes1::1lt ef a fiFe saR ee Feleaseel te tl=le eR>1iFeRmeRt via eRe ef tl=le felle¥.1iR§ patl=li.\iays: tl=le RB eFeatl=leF l=lePA filteF, tl=le RB veRtilatieR p1::1F§e system l=lePA filteFs, eF 1::1RfilteFeel leaka§e from tl=le RB. AR a1::1tematis iselatieR valve iR tl=le eFeatl=leF liRe 1::1pstFeam ef tl=le l=lePi-~, filteF is elesi§Reel te slese at (:).25 psi RB eveFpFess1::1i:e. If tl=le fiFe elees Ret Fes1::1lt iR aR RB e>1eFpFess1::1Fe §FeateF tl=laR Q.25 psi tl=le Felease 1.¥e1::1lel ee tl=IFel::l§R tl=le QQ% effisieRt l=lePA filteF iR tl=le RB BFeatl=leF Hfle. If tl=le RB >1eRtilatieR aRel p1::1F§e system is eperatiR§ aRel fails te iselate eR a Q.25 psi pFess1::1Fe si§Ral, tl=le Felease we1::1lel ee tl=IFe1::1§l=I tl=le QQ% effisieRt l=lePA filteFS iR tl=le RB P1::1F§e System exl=la1::1st liRe. If tl=le RB P1::1F§e System is Ret epeFatiR§ aRel tl=le a1::1tematis iselatieR val>.1e iR tl=le RB eFeatl=leF liRe sleses, tl=le RB i.ve1::1lel ee effestively iselateel witl=I aRy Felease eeiR§ tl=IFel::l§R aR 1::1RfilteFeel leal~a§e patl=I. IA all tl=le aeeve sseRaFies, eRly ~ % eF less ef tl=le RB aiFeeme astivity we1::1lel ee mleaseel te tl=le eR>1iroRmeRt. Ne majeF elesemmissieRiR§ astivities 1.vill ess1::1F el1::1FiR§ Pl=lase ~a. +l=le Pl=lase ~ a seRelitieR is a seRtiR1::1atieR ef tl=le P9MS seRelitieR. Wl=lile iR Pl=lase ~a, +Ml 2 Sel1:1tieRs sem~lies witl=I tl=le P9MS +S as Feviseel ey tl=le iss1::1aRse ef tl=le LiseRso +FaRsfeF AmoRelmeRt wl=lisl=I was approvoel ey tl=le NRG iR RefemRse 23. Following Phase 1a, TMl-2 will enter Phase 1b and Phase

2. During Phase 1b and Phase 2, major decommissioning operations as defined in 10 CFR 50.2 will be performed.

Limiting Condition for PDMS 3.1.1.2 and associated Surveillance Requirement 4.1.1.2 are proposed for deletion upon entry into Phase 1b and Phase 2. The purpose of the breather during PDMS is to provide passive pressure control of the containment relative to atmospheric pressure and to support measurement of unfiltered leakage from containment. With the construction of an engineered opening in containment at the RB 10of37

Attachment 5 TMI2-RA-COR-2023-002 Current TMl-2 TS Basis for Deletion 3.1.1.2 Unfiltered Leak equipment hatch, the breather no longer provides a Rate Testing preferred path to the atmosphere and an unfiltered leakage test is not required. As described in Reference 4 during Phase 1 b the RB is a passive radiological barrier to the extent that the door between the RB and the outside structure would only be open for the period of time necessary to allow passage of material or personnel between the two structures. As Elisel:!sseEI iR SeGtieR 2, aR aRalysis Ras eeeR 13eFfeFmeEI ef a l=IIG fiFe eGG!:!FFiRg el:!tsiEle ef eeRtaiRmeRt, wRiGR Fe13FeseRts tRe mast limitiRg aeeiEleRt te eGGl:!F iR PRase ~ e aREI PRase 2 Felative te Elese at tl=te site ee!:!RElaPJ. +l=te Fes!:!lts ef tRe aRalysis veFify tRe 13laRt eAB aREI bP.l sitiRg's meet tRe ~ Q GJ;:R ~ QQ.~ ~ Elese GFiteFia, as \vell as tRe ePA PAG GFiteFia; eele1.v wRiGR eff site emeF§eRey Fes13eRse ea13aeilities aFe Ret FeE11:JiFeEI. +Re aRalysis fuFtReF ElemeRstr:ates tRat tRe r:aElielegieal eeRseEjl:!SRGes ef tRe l=IIG fiFe ee!:!REI tRe Elese at tRe site ee!:!RElaPJ El!:!e te fiFe iRsiEle eeRtaiRmeRt witR tRe RB veRtilatieR aREI f31:!F§e system iR e13er:atieR. +Re Elese aRalysis asseeiateEI v.iitR tRe l=IIG firn Elees Ret take GFeElit feF mitigatiRg SSG's. During normal operation, any air flow would be into containment due to operation of the RB Purge Exhaust System. In the unlikely event that a fire occurs while this interface door is open the FPPE will direct actions to be taken. In general this will include closing the interface door by personnel involved in the material transfer and/or entry into or exit from the RB, but there may be instances when this is not the correct action. Additionally, as the RB Exhaust Ventilation would be operational at that time, the release would be through a filtered pathway. If RB Exhaust Ventilation is secured during repairs or maintenance, the interface door will generally be closed unless allowed open under controls specified in Radiation Protection and Fire Protection Programs. Additionally, work in the RB may be allowed on a case-by-case basis as determined by RP Management and consistent with the TMl-2 Fire Protection Program. Analyses were performed as described in the response to RAl-3 from Reference 3 provided in Attachment 1. 11 of 37

Attachment 5 TMI2-RA-COR-2023-002 Current TMl-2 TS Basis for Deletion 3.1.1.2 Unfiltered Leak These analyses established activity and dose rate limits to Rate Testing maintain the consequences of an off-site release below an Emergency Action Level (EAL) as defined by NEI 99-01 [Reference 7] for permanently defueled reactors (i.e., Notification of an Unusual Event (NOUE) and Alert). Guidance for implementation of administrative controls to meet the limits was also provided. Thus, maintaining the ability to perform unfiltered leak rate testing is not required in Phase 1b and Phase 2. Therefore TS 3.1.1.2 and associated Surveillance Requirement 4.1.1.2 are proposed for deletion.Thus, maintaining the ability to perform unfiltered leak rate testing is not required in Phase 1band Phase 2. Therefore TS 3.1.1.2 and associated Surveillance Requirement 4.1.1.2 are proposed for deletion. 12 of37

Attachment 5 TMI2-RA-COR-2023-002 Current TMl-2 TS Basis for Deletion 3.1.1.3 Containment Air This Technical Specification is proposed for deletion. Locks This Technical Specification assures that the containment airlocks remain operable to provide containment isolation. Containment isolation is maintained as a contamination barrier for the residual contamination which resides inside the containment. The RPP and its associated implementing procedures will maintain control of residual contamination inside containment. The airlocks are designed as a double-door system with one of the doors always closed during routine entry into the containment. There are situations when it is necessary to open both doors of an air lock assembly simultaneously such as removing large items from the containment or providing access to containment for large machines or components. Typically, both airlock doors would be open only for the period of time necessary to complete the relevant activity. The RPP and associated implementing procedures will specify the requirements necessary to perform this evolution to ensure occupational dose remains ALARA and below the occupational dose limits in 10 CFR Part 20 during Phase 1a. t>Je Ffla:ier deseFf!Ff!issiening astivities will essur during Phase 1a. +he Phase 1a senditien is a sentinuatien ef the PQMS senditien. !,OJhile in Phase 1a, +Ml 2 Selutiens seFfl13lies 1.vith the PQMS +S as revised sy the issuanse ef the bisense 

                       +ransfer /\Ff!endFflent whish was a1313reved sy the NRG in Reference 23.

Following Phase 1a, TMl-2 will enter Phase 1b and Phase 2. During Phase 1b and Phase 2, major decommissioning activities operations as defined in 10 CFR 50.2 will be performed. In Phase 1band Phase 2 procedures will be employed to control containment access. Airborne radiation monitoring will be provided to monitor containment openings for airborne radioactivity. The containment airlocks used as a means of containment access will no longer be credited for maintaining containment isolation. 

                       /\s dissussed in Sestien 2, an analysis has seen 13erferR=1ed ef a 1=41C fire essurring eutside ef sentainFflent, 1.vhish re13resents the Ff!est liFfliting assident te essur in Phase 1s relative te dese at the site seundary. +he results ef the analysis verify the 13lant 13 of 37

Attachment 5 TMl2-RA-COR-2023-002 Current TMl-2 TS Basis for Deletion 3.1.1.3 Containment e1A,B anEI bP:6 meet tl=te ~ Q CJ;:R ~ QQA ~ Elese sFiteFia as 11.iell Air Locks astl=te eP,6. PAG sFiteFia; eelew wl=tiel=t e:ff site emeF§eney Fes13ense sa13aeilities aFe net FeE11:1iFeEI. +l=te analysis fuFtl=teF Elemenstrates tl=tat tl=te raElielegisal senseEj1:1enses ef tl=te l=UC fiFe ee1:1AEI tl=te Elese at tl=te site be1:1nElai:y E11:1e te fiFe iAsiEle sentainment 1;vitl=t tl=te RB ventilatien anEI 131:1Fge system in e13eratieA. +l=te Elese aAalysis assesiateEI 1i\1itl=t tl=te l=IIC fiFe Elees not take sFedit fer mitigating SSC's. 

                   +l=te Elese assesiateEI i.¥itl=t tl=te 1.iel1:1me ef seAtaiAmeAt atmes131=tere releaseEI te tl=te envirens via airlesk e13enings in a

Pl=tase~ witl=t er witRe1:1t RB 1.ientilatien anEI 131:1rge system e13eratiAg is signifisantly less tl=tan tl=te Elese assesiateEI witl=t tl=te l=IIC fire. +Rerefere, it is sensl1:1EleEI tRat tRe res1:1lts ef tl=te aAalysis fer tl=te l=IIC fire e1:1tsiEle ef seAtainmeAt be1:1AEls any release frem an e13en airlesk. Analyses were performed as described in the response to RAI-3 from Reference 3 provided in Attachment 1. These analyses established activity and dose rate limits to maintain the consequences of an off-site release below an Emergency Action Level (EAL) as defined by NEI 99-01 [Reference 7] for permanently defueled reactors (i.e., notification of an Unusual Event (NOUE) and Alert). Guidance for implementation of administrative controls to meet the limits was also provided. The opening and closing of the containment airlocks will be controlled by the RPP. The RPP will identify the controls that will be implemented through procedures during D&D activities occurring inside of the RB. Implementation of these procedures takes into account detailed work planning, and execution of the D&D work and support activities, as well as measures to maintain occupational dose As Low As Reasonably Achievable (ALARA) and below the occupational dose limits in 10 CFR Part 20 during decommissioning. Limiting Condition for PDMS 3.1.1.3 and associated Surveillance Requirement 4.1.1.3 are proposed for deletion upon entry into Phase 1 b and Phase 2. The RPP and associated implementing procedures will provide control of residual contamination at the airlock. Procedures are utilized to control routine containment access via the air lock openings. 

                   ,Airborne radiation monitoring at the containment openings is provided. +l=te Elese at tl=te site 0e1:1nElary frem tl=te HIC firebe1:1AEls any release te tl=te em1ireAmeAt frem aA 0130A
                   .airlesk. Therefore, TS 3.1.1.3 and associated surveillance requirement 4.1.1.3 are no longer required.

14of37

Attachment 5 TMI2-RA-COR-2023-002 TS SECTION 3/4.2-REACTOR VESSEL FUEL TS Section 3/4.2 contains Limiting Conditions for PDMS to assure that no more than 42 kg of fuel (i.e., UO2) may be removed from the Reactor Vessel (RV) without prior NRC approval and that no more than 42 kg of fuel in the RV may be rearranged outside the geometries analyzed in the Defueling Completion Report (Reference 8) and the criticality safety analyses as discussed in Reference 9. The Safe Fuel Mass Limit (SFML) in the Reactor Vessel (RV) was determined to be 93 kg of core debris. Based on past industry practice, a limit of approximately 45% of the SFML was placed on the amount of Fuel Bearing Material that may be removed from the RV or rearranged in the RV. This limit is specified to ensure subcriticality even after dual errors. Calculation TMl2 EN RPT 0001 "Determination of the Safe Fuel Mass Limit for Decommissioning TMI 2,"*Nhich is presented as Attachment 5 Reference 10 provides the basis to increase the SFML from 42 kg to 1200 kg. This calculation demonstrates that the remaining core Fuel Bearing Material cannot be configured into an arrangement whereby a criticality event is possible. TS section 3/4.2 is proposed for deletion in its entirety. This TS does not apply to TMl-2 while in Phase 1b and Phase 2. Current TMl-2 TS Basis for Deletion 3/4.2-Reactor Vessel Calc1:1lation TMl2 EN RPT 0001 "Determination of the Safe Fuel Fuel Mass bimit for Decommissioning TMI 2," Reference 10 provides the basis to increase the SFML from 42 kg to 1200 kg. The results of this calculation demonstrate that the core debris cannot be configured into an arrangement whereby a criticality event is possible. A copy of the calculation is presented in Attachment 5 Reference 10. Development of the SFML includes optimization of parameters for maximum moderation, reflection and interaction conditions in a bounding geometry. The derived SFML bounds the entire expected fissile mass inventory throughout all physically separated areas within the reactor building. It is not logistically possible for the entire mass of the RV, which is deposited in a large area, to be removed at one time. Nor is it possible for the entire mass associated with the RV to be placed into a single transportable storage container (TSC). The nature of segmentation operations separate and reduce the amount of fissile material in a single area and subsequently into any TSC. It is estimated that approximately 12-14 TSCs are necessary to pack the reactor components and internals when they are packed efficiently. Procedures will be implemented to prevent stockpiling of core debris removed from the RV in one location prior to placement into a TSC. 15 of 37

Attachment 5 TMI2-RA-COR-2023-002 Current TMl-2 TS Basis for Deletion 3/4.2-Reactor Vessel The above discussion demonstrates that it is not credible Fuel for all the remaining fuel material to be collocated and for the resulting mass to exceed the determined 1200 kg U (1361 kg UO2) SFML. It is further not considered credible that the segmentation operations would result in the fuel material being placed in an optimal physical arrangement that maximizes reactivity. All credible operational upset conditions associated with the remaining fuel in the facility are bounded such that a criticality accident during decommissioning operations is not credible. Therefore, TS 3/4.2 is proposed for deletion. 16 of37

Attachment 5 TMI2-RA-COR-2023-002 TS SECTION 3/4.3-CRANE OPERATIONS TS Section 3/4.3 contains a Limiting Condition for PDMS to assure that loads in excess of 50,000 lbs. are prohibited from travel over the Reactor Vessel (RV) unless a docketed safety evaluation for the activity is approved by the NRC. This is to preclude a load drop into the RV that may cause reconfiguration of the core debris outside the analyzed geometries used in the "Defueling Completion Report" (Reference 8) and the criticality safety analysis as discussed in Reference 9. TS section 3/4.3 is proposed for deletion in its entirety. Current TMl-2 TS Basis for Deletion 3/4.3 Crane Operations This Technical Specification is proposed for deletion. No major decommissioning will occur during Phase 1a. The Phase 1a condition is a continuation of the PDMS condition. VVhile in Phase 1a, TMI 2 Solutions complies with the PDMS TS as revised by the issuance of the Lioense Transfer Amendment which was approved by the NRG in Reference 23. Following Phase 1a, TMl-2 will enter Phase 1b and Phase 2. During Phase 1b and Phase 2, major decommissioning activities as defined in 10 CFR 50.2 will be performed. The PDMS TS requirements associated with TS 3/4.3 "Crane Operations," are not applicable in Phase 1b and Phase 2. TS 3/4.3 does not satisfy any of the four requirements established in 10 CFR 50.36(c)(2)(ii) based upon the evaluation provided in Section 3.1 "Applicable Regulatory Requirements". Criterion 1 10 CFR 50.36( c)(2)(ii)(A) states that TS limiting conditions for operation must be established for "installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary." TMl-2 does not have a reactor coolant pressure boundary; therefore, the requirements of Criterion 1 of 10 CFR 50.36(c)(2)(ii)(A) are not applicable. 17 of37

Attachment 5 TMI2-RA-COR-2023-002 Criterion 2 10 CFR 50.36(c)(2)(ii)(B) states that TS limiting conditions or operation must be established for a "process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." TMl-2 is no longer licensed to operate, therefore the requirements of Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B) are not applicable. Criterion 3 he requirements of Criterion 3 of 10 CFR 50.36( c) (2)(ii)(C) states that Technical Specification limiting conditions for operation must be established for "A SSC that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The TMl-2 cranes do not provide a function required to mitigate the effect of unanticipated occurrences such as the ire in containment as described in the response to RAl-3 rom Reference 3 provided in Attachment 1, ,thus the requirements of Criterion 3 of 10 CFR 50.36(c)(2)(ii)(C) are not applicable. Criterion 4 he requirements of Criterion 4 of 10 CFR 50.36(c)(2)(ii)(D) states that TS limiting conditions for operation must be established for "A SSC which operating experience or probabilistic risk assessment has shown to be significant to public health and safety." The intent of this criterion is that risk insights and operating experience be factored into the establishment of TS LCOs. here are no TS associated with Phase 1b or Phase 2, hence there are no limiting conditions for operation. However, TMl-2 has a procedure that provides a methodology for gathering OP EX/LL information from various sources for systematic review, evaluation, and use on the TM 1-2 project. 18 of 37

Attachment 5 TMI2-RA-COR-2023-002 ~s the analyses performed as described in the response to RAl-3 from Reference 3 provided in Attachment 1. do not credit containment closure and limits a release to the environment to below the requirements for an NOUE, the requirements of Criterion 4 of 10 CFR 50.36(c)(2)(ii)(D) are not applicable. rro provide a high degree of assurance that a load drop into ~he reactor vessel will not occur TMl-2 Solutions wiU !~~*:=~~~ has developed a hoisting and rigging program that addresses movement of loads at TMl-2. The purpose of the hoisting and rigging program is to define the minimum requirements for the safe operations of cranes and hoists. The hoisting and rigging program will provide provides detailed requirements for training and qualification of personnel, inspection and maintenance of cranes or hoists, the safe use of rigging equipment as well as documentation for Non-Standard Lifts. A Non-Standard Lift has characteristics that require additional planning and performance efforts to ensure that the lift is performed in a safe manner. A lift plan will be developed for all Non-Standard lifts as directed by the hoisting and rigging program. Implementation of the hoisting and rigging program provides a defense in depth approach to preventing a load drop from occurring. Crane design features such as load cells, and travel stops, will be employed as required in order to ensure safe travel paths. Steel plates and barriers will be provided as required by the lift plan to preclude the effects of a load drop. In addition to the above, a calculation (Attachment 5) (Reference 10) has been performed that assesses increasing the Safe Fuel Mass Limit (SFML) from 42 kg to approximately 1200 kg. All credible operational upset conditions associated with the remaining fuel in the facility are bounded such that a criticality accident during decommissioning operations is not credible. The attributes of the hoisting and rigging program, and lift plan, coupled with the results of the criticality analysis provided in Attachment 5 Reference 10, ensure that crane operations at TMl-2 can be performed safely. Therefore, Technical Specification 3/4.3 is proposed for deletion. 19 of37

Attachment 5 TMI2-RA-COR-2023-002 TS SECTION 3/4.4-SEALED SOURCES TS Section 3/4.4 contains Limiting Condition for PDMS to assure that each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material (except as noted in surveillance requirement 4.4.1.2) are free of ;::0.005 microcuries of removable contamination. The limitation on removable contamination for sources requiring leak testing, including alpha emitters, is based or:, 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and Special Nuclear Material sources will not exceed allowable values. The TMl-2 sealed sources are maintained at TMI 1 and managed by Exelon under a program compliant with the requirements of 10 CFR 70.39(c). TS section 3/4.4 is proposed for relocation to the DQAP DSAR. The change in relocation was noted in Reference 12 Current TMl-2 TS Basis for Deletion 3/4.4 Sealed Sources This Technical Specification is proposed for deletion and transfer to the TMl-2 DSAR.Section 50.36 of 10 CFR provides four criteria to define the scope of equipment and parameters to be included in the TS. Each of these criteria are addressed below. TS 3/4.4 does not satisfy any of the four requirements established in 10 CFR 50.36(c)(2)(ii) based upon the evaluation provided in Section 3.1. Criterion 1 10 CFR 50.36(c)(2)(ii)(A) states that TS limiting conditions for operation must be established for "installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary." TMl-2 does not have a reactor coolant pressure boundary; therefore, the requirements of Criterion 1 of 10 CFR 50.36( c)(2)(ii)(A) are not applicable. Criterion 2 10 CFR 50.36(c)(2)(ii)(B) states that TS limiting conditions for operation must be established for a "process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." TMl-2 is no longer licensed to operate, therefore the requirements of Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B) are not applicable. 20 of 37

Attachment 5 TMI2-RA-COR-2023-002 Current TMl-2 TS Basis for Deletion Criterion 3 The requirements of Criterion 3 of 10 CFR 50.36(c)(2)(ii)(C) states that Technical Specification limiting conditions for operation must be established for "A SSC that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." Sealed sources do not provide a function required to mitigate the effect of unanticipated occurrences such as the fire in the RB as described in the response to RAl-3 from Reference 3 provided in Attachment 1 . Thus, the requirements of 10 CFR 50.36( c)(2)(ii)(C) do not apply. Criterion 4 The requirements of Criterion 4 of 10 CFR 50.36(c)(2)(ii)(D) states that TS limiting conditions for operation must be established for "A SSC which operating experience or probabilistic risk assessment has shown to be significant to public health and safety." The intent of this criterion is that risk insights and operating experience be factored into the establishment of TS LCOs. There are no TS associated with Phase 1b or Phase 2, hence there are no limiting conditions for PDMS. However, TMl-2 has a procedure that provides a methodology for gathering OPEX/LL information from various sources for systematic review, evaluation, and use on the TMl-2 project. Sinoe the oontainment fire analysis does not oredit oontainment olosure and assumes a release to the environment, whioh is bounded by the results of the HIC fire analysis, the requirements of Criterion 4 of 10 CFR 50.36(o)(2)(ii)(D) are not applioable. The analyses performed as described in the response to RAl-3 from Reference 3 provided in Attachment 1 do not credit containment closure and limits a release to the environment to below the requirements for an NOUE,, the requirements of Criterion 4 of 10 CFR 50.36(c)(2)(ii)(D) are not applicable. 21 of 37

Attachment 5 TMI2-RA-COR-2023-002 Current TMl-2 TS Basis for Deletion Therefore, the requirements of Technical Specification 3/4.4 are deleted from the TS and controlled under a program which meets the requirements of 10 CFR 70.39(c) and will be transferred to the DSAR and thus be subject to 10 CFR 50.59 change evaluation process that ensures adequate regulatory controls are in place. 22 of 37

Attachment 5 TMI2-RA-COR-2023-002 TS SECTION 5.0 DESIGN FEATURES TS Section 5.0 "Design Features," contains design parameters related to the containment configuration. This information is being proposed for deletion. Current TMl-2 TS Proposed TMl-2 TS 5.1 CONTAINMENT This section is proposed for deletion in its entirety. CONFIGURATION Basis TS Section 5.1 "Containment, Configuration" identifies principal design parameters associated with a robust containment structure designed to accommodate pressurization. During PDMS, containment isolation is maintained to assure the containment is properly maintained as a contamination barrier for the residual contamination which resides inside the containment. Phase 1a is a continuation of the PDMS condition. Following Phase 1a, TMl-2 will enter Phase 1b and Phase 2. During Phase 1band Phase 2, major decommissioning activities as defined in 10 CFR 50.2 will be performed. Control of residual contamination inside of containment will be provided by the Radiation Protection Program (RPP) and through implementation of procedures which address execution of D&D work and support activities, personnel safety as well as measures to maintain occupational dose ALARA and below the occupational dose limits in 10 CFR Part 20 during decommissioning as applicable. In support of D&D, engineered access openings in the RB basement and at the RB equipment hatch will be constructed. Control of residual contamination at the containment access openings will be provided by the RPP and implementing procedures. Airborne radiation monitoring will be provided at the containment openings. J;:uFtheF meFe,the dese at tl=le site beundaFy asseciated 11.iitl=I tl=le l=IIG fiFeas EliscusseEI in Sectien 2"QetaileEI QescFiptien AnEI Basis F9F +l=le Gl=lan~es,"dees net exoeed the rnquirnments sf 10 GJ;:R 100.11 and the EPA PAGs. Furthermore, As described in the response to RAl-3 from Reference 3 provided in Attachment 1. analyses have been performed that demonstrate that there are no postulated unanticipated events that could occur during Phase 1b and Phase 2, that would result in accident releases exceeding the requirements of 10 CFR 100.11 and the EPA PAGs. 10 CFR 50.36 describes the design features to be included in Technical Specifications, as those features of the facility, such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety. As described above there are no postulated unanticipated events that could occur during Phase 1b and Phase 2, that would result in accident releases exceeding the requirements of 10 CFR 100.11 and the EPA PAGs. AEIElitienally, tl=lis same infeFmation oan be feunEI in tl=le PQMS SAR. Therefore, the containment design features presented in TS Section 5.0 are proposed for deletion. 23 of37

Attachment 5 TMI2-RA-COR-2023-002 TS SECTION 6.0 ADMINISTRATIVE CONTROLS The existing TS Section 6.0 Administrative Controls contains provisions relating to organization and management, procedures, recordkeeping, review and audit, programs, and reporting necessary to assure operation of the facility in a safe manner. NRC Administrative Letter 95-06 (Reference 11) provides a discussion concerning the relocation of Technical Specification administrative controls to a Quality Assurance (QA) program. The NRC considers relocating certain TS Administrative Controls to a licensee QA program acceptable because of the controls imposed by Appendix B to 10 CFR Part 50, the existence of an NRC approved QA program, and the quality assurance program change control process as presented in 10 CFR 50.54(a). After these administrative controls are incorporated into the TMl-2 DQAP, any future changes are controlled in accordance with 10 CFR 50.54(a). Several of the TS Section 6.0 Administrative Controls are no longer applicable in Phase 1band Phase 2 and are being proposed for deletion. Only those TS Section 6 Administrative Controls that have changed from Reference 1 are listed Current TMl-2 TS Proposed TMl-2 TS Relocated to DQAP 6.7 PROCEDURES AND PROGRAMS 6.7 PROCEDURES AND PROGRAMS 6.7.1 Written procedures shall be 6.7.1 Written procedures shall be established, implemented, and established, implemented, and maintained for the activities maintained for the activities to be necessary to maintain the PDMS performed in Phase 1b and condition as described in the PDMS Phase 2 as described in the SAR. Examples of these activities DSAR. necessary to maintain the are: PDMS condition as described in the

a. Technical Specification PDMS SAR. Examples of these implementation. activities are:
a. Technical Specification
b. Radioactive waste management and shipment. implementation.
c. Radiation Protection Plan b. Radioactive waste management Implementation. and shipment.
d. Fire Protection Program c. Radiation Protection Plan implementation. Implementation.
e. Flood Protection Program d. Fire Protection Program implementation.

implementation.

e. Flood Protection Program implementation.

Basis As described in Reference 12 Proposed Technical Specification 6. 7 .1 contained an administrative error in that the words" ... as described in the ... " were inadvertently deleted thus changing the meaning of this Technical Specification. 24 of37

Attachment 5 TMI2-RA-COR-2023-002 Current TMl-2 TS Proposed TM 1-2 TS 6.8 REPORTING REQUIREMENTS 6.8 REPORTING REQUIREMENTS ROUTINE REPORTS ROUTINE REPORTS ANNUAL RADIOACTIVE EFFLUENT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT RELEASE REPORT 6.8.1.2 The Annual Radiological Effluent 6.8.1.2 The Annual Radiological Effluent Release Report covering the operation of Release Report covering the operation of the unit during the previous calendar year the -t::ffi#-facility during the previous shall be submitted before May 1 each year. calendar year shall be submitted before The report shall include a summary of the May 1 each year. The report shall include quantities of radioactive liquid and gaseous a summary of the quantities of radioactive effluents and solid waste released from liquid and gaseous effluents and solid the unit. The material provided shall be (1) waste released from the +Jffitfacility. The consistent with the objectives outlined in material provided shall be (1) consistent the ODCM and (2) in conformance with 10 with the objectives outlined in the ODCM CFR 50.36a and Section IV.B.1 of and (2) in conformance with 10 CFR Appendix I to 10 CFR Part 50. 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50. Basis 25 of 37

Attachment 5 TMI2-RA-COR-2023-002 The proposed changes described above are reflected in Attachments 6 and 7. This NSHC revision supersedes the NSHC provided in the letter dated January 7, 2022 (ML22013A177). Change bars indicate revisions to the NSHC provided in that letter. Revisions made to the NSHC analysis do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c). The NSHC submitted in the letter dated January 7, 2022 (ML22013A177) remains bounding. Revised No Significant Hazards Consideration (NSHC) Pursuant to 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," TMl-2 Solutions LLC, proposes an amendment to the Possession Only License (POL) and Technical Specifications, of POL No. DPR-73 for Three Mile Island Nuclear Station, Unit 2 ("TMl-2"). This proposed LAR, upon approval, will revise the POL and the associated TS to support the transition of TMl-2 from PDMS to that of a facility undergoing decommissioning. The proposed amendment would revise the POL and TS to support Phase 1b and Phase 2 activities associated with achieving the removal of all Debris Material Fuel Bearing Material, its transfer to dry cask storage at an Independent Spent Fuel Storage Installation (ISFSI), or to a suitable waste storage area, and the relocation of various requirements to the TMl-2 DQAP and Sealed Sources to the Defueled Safety Analysis Report (DSAR). As noted in a letter from the NRG to GPU Nuclear dated February 13, 2013, (Reference

1) the equivalent to the certificate of cessation of operations was determined to be the NRC's issuance of TMl-2 License Amendment 45, converting the TMl-2 operating license to a possession only license. This amendment was granted on September 14, 1993 (Reference 13) and establishes that date as the date that TMl-2 is considered to have submitted certification of permanent cessation of operations.

The proposed changes to the POL and TS, for deletion or revision, are in accordance with 10 CFR 50.36(c)(1) through 10 CFR 50.36(c)(5). The proposed changes also include a renumbering of pages and sections, where appropriate, to condense and reduce the number of pages in the TS without affecting the technical content. The existing TMl-2 TS contain Limiting Conditions for PDMS that provides the functional capability of equipment required for safe operation of the facility. The current TS are only applicable with TMl-2 in the PDMS condition. Limiting Conditions for PDMS and associated Surveillance Requirements (SRs) that will not apply in Phase 1b or Phase 2 are being proposed for deletion. The remaining portions of the TS are being proposed for revision and will continue to provide an acceptable level of control for the TMl-2 facility as it undergoes decommissioning. TMl-2 Solutions has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 1o-CFR 50.92, "Issuance of amendment," as discussed below:

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

26 of 37

Attachment 5 TMI2-RA-COR-2023-002 Response: No. The proposed changes would revise the TMl-2 POL and TS by deleting or modifying certain portions of the TS that are no longer applicable to TMl-2 as it transitions from PDMS to decommissioning. This change is consistent with the criteria set forth in 10 CFR 50.36 for the contents of TS. The Phase 1a condition is a continuation of the PDMS condition. l'Jo ma;ior decommissioning activities will occur in Phase 1a. As discussed in Section 2 "Detailed Description and Basis for the Changes" of this proposed amendment, the radiological consequences associated with the fire inside containment unanticipated event does not exceed the applicable limits of 10 CFR 100.11 and the EPA PAGs. Following Phase 1a, TMl-2 will enter Phase 1b and Phase 2. During Phase 1band Phase 2, major decommissioning activities as defined in 10 CFR 50.2 will be performed. As discussed in Attachment 1 to this letter, none of the events evaluated exceed the applicable limits of 10 CFR 100.11 and the EPA PAGs. The response to RAl-3 from Reference 3 provided in Attachment 1 describes the events associated with TMl-2 decommissioning. For the events analyzed the dose at the boundary remains within the limits of 10 CFR 100.11 and the EPA PAGs. During Phase 1a, containment isolation is maintained to assure the containment is properly maintained as a contamination barrier for the residual contamination which resides inside the containment. There are no postulated accidents that can occur inside of the Reactor Building (RB) during Phase 1b or Phase 2 that result in the dose at the site boundary exceeding the limits of 10 CFR 100.11 and the EPA PAGs including such times as when the containment engineered access equipment hatch is open. The D&D process includes many evolutions that will require the equipment hatch and other RB access points to be open to allow movement of equipment, waste, and other materials into and out of the RB. The Radiation Protection Plan (RPP) will identify the controls that will be implemented through procedures during D&D activities occurring inside of the RB. Implementation of these procedures take into account detailed work planning, and execution of the D&D work and support activities, including measures to maintain occupational dose As Low As Reasonably Achievable (ALARA) and below the occupational dose limits in 10 CFR Part 20 during decommissioning. As described in Reference 4, credit is being taken for the Reactor Building as a passive radiological barrier to the extent that the door between the Reactor Building and the outside structure would only be open for the period of time necessary to allow passage of material or personnel between the two structures. During normal operation, any air flow would be into containment due to operation of the Reactor Building Purge Exhaust System. In the unlikely event that a fire occurs while this interface door is open it will be closed by personnel involved in the material transfer and/or entry into or exit from the Reactor Building. Additionally, as the Reactor Building Exhaust Ventilation would be operational at that time, the release would be through a filtered pathway. If Reactor Building Exhaust Ventilation is secured during repairs or maintenance, the interface door will be closed. Work in the Reactor Building may be allowed on a case-27 of 37

Attachment 5 TMI2-RA-COR-2023-002 by-case basis as determined by RP Management and consistent with the TMl-2 Fire Protection Program. Procedures associated with Phase 1b will be developed to retrieve the remaining core debris and decontaminate high radiation areas. Phase 2 procedures will also be developed; however, the focus of these procedures is related to performing D&D operations in a facility which has not experienced an accident. The deletion of TS 3/4.1 does not cause a change in facility conditions, design function, or analysis that verifies the ability of SSCs to perform a design function. The function of the containment is to contain residual contamination inside the containment during decommissioning remains unchanged. During Phase 1b the RPP and associated implementing procedures will provide the controls necessary to manage residual contamination. As such the containment continues to function as a contamination barrier. Airborne radiation monitoring will be provided at the engineered containment openings (e.g., Equipment Hatch Opening). With the construction of the engineered openings in containment the RB breather no longer provides a preferred path to the atmosphere. No credit is taken for the containment as a pressure containing boundary and therefore unfiltered leak rate testing of the containment is no longer applicable. For the events analyzed the dose at the boundary remains within the limits of 10 CFR 100.11 and the EPA PAGs. Therefore, the deletion of TS 3/4.1 "Containment" does not involve a significant increase in the probability or consequences of an accident previously evaluated. TS 3/4.2 "Reactor Vessel Fuel" establishes a Safe Fuel Mass Limit (SFML) for the PDMS condition, which ensures that the amount of core debris that may be removed from the Reactor Vessel (RV) or rearranged in the RV during PDMS does not exceed 42kg. This SFML is specified to ensure subcriticality even after dual errors. The deletion of TS 3/4.2 does not cause a change in facility conditions, design function, or analysis that verifies the ability of SSCs to perform a design function. A calculation is presented in Attachment 5 (of Reference 1) which Reference 10 provides the basis to increase the SFML from 42 kg to 1200 kg. The result of this calculation demonstrates that the entire mass of the core debris material cannot be configured into an arrangement whereby a criticality event is possible and that Kett does not exceed 0.95. Therefore, the deletion of TS 3/4.2 "Reactor Vessel Fuel" does not involve a significant increase in the probability or consequences of an accident previously evaluated. In Phase 1a, TS 3/4.3 "Crane Operations" prohibits loads in excess of 50,000 lbs. from travel over the RV. The deletion of TS 3/4.3 does not cause a change in facility conditions, design function, or analysis that verifies the ability of SSCs to perform a design function. As discussed in Section 2 "Detailed Description and Basis for The Changes," for Phase 1b and Phase 2, TMl-2 Solutions will develop has developed a hoisting and rigging program that addresses movement of loads at TMl-2. The purpose of the hoisting and 28 of37

Attachment 5 TMI2-RA-COR-2023-002 rigging program is to define the minimum requirements for the safe operations of cranes and hoists. The hoisting and rigging program will provide provides as applicable, detailed requirements for training and qualification of personnel, inspection and maintenance of cranes or hoists, the safe use of rigging equipment as well as direction for performing Non-Standard Lifts in order to ensure that lifting operations are performed in a safe manner. A lift plan will be developed for all lifts as directed by the hoisting and rigging program where a load drop or load impingement could contribute to release or dispersal of radioactive material to the environment which could exceed the threshold for an unusual event. Implementation of the hoisting and rigging program provides a defense in depth approach to preventing a load drop from occurring. Crane design features such as load cells, and travel stops, will be employed as required to ensure safe travel paths. Barriers will be provided as per the lift plan, as required to preclude the effects of a load drop. Based on the above the PDMS Quality Assurance Program for TMl-2 will also be modified from: Lifting and Handling activities, including testing and surveillance of cranes and rigging components where the equipment and activities involve the handling or movement of radioactive material where a load drop or load impingement could contribute to unplanned release or dispersal of radioactive material or where such activity involves the movement of loads over the Reactor Vessel, or the handling of material that could contain Special Nuclear Material. To: Lifting and Handling activities for all lifts where a load drop or load impingement 

      . could contribute to release or dispersal of radioactive material to the environment which could exceed the threshold for an unusual event.

A calculation has been performed (Attaohment 5 of Referenoe 1 Reference 10) that assesses increasing the Safe Fuel Mass Limit (SFML) from 42 kg to approximately 1200 kg. The analysis states that it is not credible to have 1200 kg U in an idealized configuration for criticality to occur. There are no credible operational upsets to realize the ideal configuration but even in the event that the upset occurs, it would require fissile mass in excess of that analyzed, which is greater than what is anticipated. Therefore, the deletion of TS 3/4.3 "Crane Operations" does not involve a significant increase in the probability or consequences of an accident previously evaluated. The TMl-2 sealed sources are maintained under at TMI 1 and managed by Exelon a program compliant with the requirements of 10 CFR 70.39(c). Deleting TS 3/4.4 "Sealed Sources" from the TMl-2 TS and relocating the TS requirements to the DSAR does not involve a significant increase in the probability or consequences of an accident previously evaluated. The deletion of TS definitions and rules of usage and application that will not be applicable during Phase 1b and Phase 2 decommissioning, has no impact on facility structures, systems, and components (SSCs) or the methods of operation of such SSCs. 29 of 37

Attachment 5 TMI2-RA-COR-2023-002 The proposed relocation of certain administrative requirements as allowed by Administrative Letter 95-06 (Reference 11) do not affect operating procedures or administrative controls that have the function of ensuring the safe management of Debris Material Fuel Bearing Material or decommissioning of the facility. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No. The proposed changes to delete and/or modify the TS does not create the possibility of a new or different kind of accident from that previously evaluated. The removal of the TS applicable in Phase 1a cannot result in different or more adverse accidents than previously evaluated because there are no new credible failure mechanisms, or accident initiators not considered in the design and licensing basis for Phase 1b. Following Phase 1a, TMl-2 will enter Phase 1b and Phase 2. During Phase 1b and Phase 2, major decommissioning activities as defined in 10 CFR 50.2 will be performed. As discussed in- Attachment-1 to this letter -these events have been described in various License Basis Documents thus a new or different kind of accident from any accident previously evaluated has not been created. During Phase 1a, containment isolation is maintained to assure the containment is properly maintained as a contamination barrier for the residual contamination which resides inside the containment. There are no postulated accidents that can occur inside of the RB during Phase 1b or Phase 2 that result in the dose at the site boundary exceeding the limits of 10 CFR 100.11 and the EPA PAGs including such times as when the containment engineered access equipment hatch is open. The D&D process includes many evolutions that will require the equipment hatch and other RB access points to be open to allow movement of equipment, waste, and other materials into and out of the RB. As described in Reference 4, credit is being taken for the Reactor Building as a passive radiological barrier to the extent that the door between the Reactor Building and the outside structure would only be open for the period of time necessary to allow passage of material or personnel between the two structures. During normal operation, any air flow would be into containment due to operation of the Reactor Building Purge Exhaust System. In the unlikely event that a fire occurs while this interface door is open it will be closed by personnel involved in the material transfer and/or entry into or exit from the Reactor Building. Additionally, as the Reactor Building Exhaust Ventilation Would be operational at that time, the release would be through a filtered pathway. If Reactor Building Exhaust Ventilation is secured during repairs or maintenance, the interface door will be closed. Work in the Reactor Building may be allowed on a case-30 of37

Attachment 5 TMI2-RA-COR-2023-002 by-case basis as determined by RP Management and consistent with the TMl-2 Fire Protection Program. The RPP will identify the controls that will be implemented through procedures during D&D activities occurring inside of the RB. Implementation of these procedures take into account detailed work planning, and execution of the D&D work and support activities, including measures to maintain occupational dose As Low As Reasonably Achievable (ALARA) and below the occupational dose limits in 10 CFR Part 20 during decommissioning. Procedures associated with Phase 1b will be developed to retrieve the remaining core debris and decontaminate high radiation areas. Phase 2 procedures will also be developed; however, the focus of these procedures is related to performing D&D operations in a facility which has not experienced an accident. The deletion of TS 3/4.1 "Containment" does not cause a change in facility conditions, nor does it cause a change in design function. The function of the containment is to maintain residual contamination during Phase 1a remains unchanged. During Phase 1b and 2, the RPP and associated implementing procedures will provide the controls necessary to manage residual contamination. As such, the containment continues to function as a contamination barrier. Airborne radiation monitoring will be provided at the engineered containment openings (e.g., Equipment Hatch Opening). Procedures are utilized to control routine containment access. With the construction of the engineered openings in containment the RB breather no longer provides a preferred path to the atmosphere. No credit is taken for the containment as a pressure containing boundary and therefore unfiltered leak rate testing of the containment is no longer applicable. The dose at the site boundary associated with the the m1ents desoribed in Attaohment 1 to this letter TMl-2 Events described in the response to RAl-3 from Reference 3 provided in Attachment 1 do not exceed the requirements of 10 CFR 100.11, as well as the EPA PAGs. Therefore, the deletion of TS 3/4.1 "Containment" does not create the possibility of a new or different kind of accident from any accident previously evaluated relative to Phase 1b or Phase 2. TS 3/4.2 "Reactor Vessel Fuel" establishes a Safe Fuel Mass Limit (SFML) for the PDMS condition, which ensures that the amount of core debris that may be removed from the RV or rearranged in the RV during PDMS does not exceed 42kg. This SFML limit is specified to ensure subcriticality even after dual errors. The deletion of the TS does not cause a change in facility conditions, nor does it cause a change in design function. A calculation is presented in Reference 10 as Attaohment 5 (of Referenoe 1), which provides the basis to increase the SFML from 42 kg to 1200 kg. The result of this calculation demonstrates that the entire mass of the core debris material cannot be configured into an arrangement whereby a criticality event is possible and that Kett does not exceed 0.95. Therefore, the deletion of TS 3/4.2 "Reactor Vessel Fuel" does not create the possibility of a new or different kind of accident from any accident previously evaluated relative to Phase 1b or Phase 2. 31 of 37

Attachment 5 TMI2-RA-COR-2023-002 As part of the PDMS condition, loads in excess of 50,000 lbs. are prohibited from travel over the RV. The deletion of TS 3/4.3 "Crane Operations" does not cause a change in facility conditions nor does it cause a change in design function. As discussed in Section 2 "Detailed Description and Basis for The Changes," for Phase 1b and Phase 2, TMl-2 Solutions will develop has developed a hoisting and rigging program that addresses movement of loads at TMl-2. The purpose of the hoisting and rigging program is to define the minimum requirements for the safe operations of cranes and hoists. The hoisting and rigging program will provide provides detailed requirements as applicable for training and qualification of personnel, inspection and maintenance of cranes or hoists, the safe use of rigging equipment as well as direction for performing Non-Standard Lifts in order to ensure that lifting operations are performed in a safe manner. A lift plan will be developed for all lifts as directed by the hoisting and rigging program where a load drop or load impingement could contribute to release or dispersal of radioactive material to the environment which could exceed threshold for an unusual event. Implementation of the hoisting and rigging program provides a defense in depth approach to preventing a load drop from occurring. Crane design features such as load cells, and travel stops, will be employed as required to ensure safe travel paths. Barriers will be provided as required to preclude the effects of a load drop. A calculation has been performed (Attachment 5 of Reference 1) (Reference 10) that assesses increasing the Safe Fuel Mass Limit (SFML) from 42 kg to approximately 1200 kg. The analysis states that it is not credible to have 1200 kg U in an idealized configuration for criticality to occur. There are no credible operational upsets to realize the ideal configuration but even in the event that the upset occurs, it would require fissile mass in excess of that analyzed, which is greater than what is anticipated, in addition to a greatly reduced impurity concentration to present a criticality hazard. Therefore, the deletion of TS 3/4.3 "Crane Operations" does not create the possibility of a new or different kind of accident from any accident previously evaluated relative to Phase 1b or Phase 2. The TMl-2 sealed sources are maintained at TMI 1 and managed by Exelon under a program compliant with the requirements of 10 CFR 70.39(c). Deleting TS 3/4.4 "Sealed Sources" from the TMl-2 TS and relocating the TS requirements to the DSAR does not create the possibility of a new or different kind of accident from any accident previously evaluated relative to Phase 1b or Phase 2. The proposed change will not create the possibility of a new or different kind of accident due to credible new failure mechanisms, malfunctions, or accident initiators not considered in the licensing bases documents. Decommissioning operations in Phase 1b and Phase 2 are bounded by the events described in Attachment 1 the response to RAl-3 from Reference 3 provided in Attachment 1 . . I - Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated relative to Phase 1b or Phase 2. 32 of37

Attachment 5 TMI2-RA-COR-2023-002

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

Response: No. The proposed changes would revise the TMl-2 POL and TS by deleting or modifying certain portions of the TS that are no longer applicable to TMl-2 as it transitions from PDMS to decommissioning. This change is consistent with the criteria set forth in 10 CFR 50.36 for the contents of TS. The Phase 1a condition is a continuation of the PDMS condition. No major decommissioning activities will occur in Phase 1a. As discussed in Section 2 "Detailed Description and Basis for the Changes" (of Reference 1) of this proposed amendment, the radiological consequences associated with the fire inside containment, unanticipated event, does not exceed the applicable limits of 10 CFR 100.11 and the EPA PAGs .. Following Phase 1a, TMl-2 will enter Phase 1b and Phase 2. During Phase 1b, major decommissioning activities as defined in 10 CFR 50.2 will be performed. Based on the consequences of the postulated events described in Attachment 1 the response to RAl-3 from Reference 3 provided in Attachment 1. , none of the events evaluated exceed the applicable limits of 10 CFR 100.11 and the EPA PAGs or involve a significant reduction in a margin of safety. During Phase 1a, containment isolation is maintained to assure the containment is properly maintained as a contamination barrier for the residual contamination which resides inside the containment. There are no postulated accidents that can occur inside of the RB during Phase 1b or Phase 2 that result in the dose at the site boundary exceeding the limits of 10 CFR 100.11 and the EPA PAGs including such times as when the containment engineered access equipment hatch is open. The D&D process includes many evolutions that will require the equipment hatch and other RB access points to be open to allow movement of equipment, waste, and other materials into and out of the RB. As described in Reference 4, credit is being taken for the Reactor Building as a passive radiological barrier to the extent that the door between the Reactor Building and the outside structure would only be open for the period of time necessary to allow passage of material or personnel between the two structures. During normal operation, any air flow would be into containment due to operation of the Reactor Building Purge Exhaust System. In the unlikely event that a fire occurs while this interface door is open it will be closed by personnel involved in the material transfer and/or entry into or exit from the Reactor Building. Additionally, as the Reactor Building Exhaust Ventilation would be operational at that time, the release would be through a filtered pathway.

  • If Reactor Building Exhaust Ventilation is secured during repairs or maintenance, the interface door will be closed. Work in the Reactor Building may be allowed on a case-by-case basis as determined by RP Management and consistent with the TMl-2 Fire Protection Program.

33 of 37

Attachment 5 TMI2-RA-COR-2023-002 The RPP will identify the controls that will be implemented through procedures during D&D activities occurring inside of the RB. Implementation of these procedures take into account detailed work planning, and execution of the D&D work and support activities, including measures to maintain occupational dose As Low As Reasonably Achievable (ALARA) and below the occupational dose limits in 10 CFR Part 20 during decommissioning. Procedures associated with Phase 1b will be developed to retrieve the remaining core debris and decontaminate high radiation areas. Phase 2 procedures will also be developed; however, the focus of these procedures is related to performing D&D operations in a facility which has not experienced an accident. The deletion of TS 3/4.1 "Containment" does not exceed or alter a design basis or safety limit. The function of the containment (i.e., to control residual contamination during Phase 1a) remains unchanged. During Phase 1b the RPP and associated implementing procedures will provide the controls necessary to manage residual contamination. As such the containment continues to function as a contamination barrier. Airborne radiation monitoring will be provided at the engineered containment openings (e.g., Equipment Hatch Opening). Procedures are utilized to control routine containment access. With the construction of the engineered openings in containment the RB breather no longer provides a preferred path to the atmosphere. No credit is taken for the containment as a pressure containing boundary and therefore unfiltered leak rate testing of the containment is no longer applicable. The dose at the site boundary associated with the events desaribed in Attaahment 1 the TMl-2 Events described in the response to RAl-3 from Reference 3 provided in Attachment 1 bound the dose at the site boundary associated with Phase 1b and Phase 2 and does not exceed the requirements of 10 CFR 100.11 and the EPA PAGs. Therefore, deletion of TS 3/4.1 "Containment" does not significantly reduce the margin of safety during Phase 1b and Phase 2. TS 3/4.2 "Reactor Vessel Fuel" establishes a Safe Fuel Mass Limit (SFML) for the PDMS condition, which ensures that the amount of core debris that may be removed from the RV or rearranged in the RV during PDMS does not exceed 42kg. This SFML limit is specified to ensure subcriticality even after dual errors. A calculation is presented as Attaahment 5 (of Referenae 1) in Reference 10 which provides the basis to increase the SFML from 42 kg to 1200 kg. The current SFML was developed based solely on credible upper bounds for input parameters as opposed to sample data or realistic conditions. The proposed revision to the SFML is based upon existing data and known conditions. These inputs are still considered to be reasonably and sufficiently conservative for their use in development of the proposed 1200 kg SFML. The derived SFML bounds the entire expected fissile mass inventory throughout all physically separated areas within the reactor building. The bounding fissile mass used to produce the SFML is assembled in idealized conditions that cannot credibly exist during decommissioning operations. Even if the expected remaining fissile mass throughout the building, including hold up in all piping and cubicles were to be brought together, a criticality is not feasible. There are no credible operational upsets to realize the ideal configuration but even in the event that 34 of37

Attachment 5 TMI2-RA-COR-2023-002 the upset occurs, it would require fissile mass in excess of that analyzed, which is greater than what is anticipated. In addition, the SFML is based on a significantly reduced impurity concentration below that demonstrated to be present. The kefffor the new SFML in the idealized static conditions does not exceed 0.95. The calculation of the new SFML states that the entire mass of the core debris material cannot be configured into an arrangement whereby a criticality event is possible. Debris Material Fuel Bearing Material removal operations will involve loading 12-14 storage casks with each cask containing less than the total SFML calculated for Phase 1b. The overall subcritical nature, namely inherent elemental constituents, of the fuel debris remaining at the TMl-2 facility today is equivalent to that associated with the fuel debris at TMl-2 prior to defueling operations. The presence of some intact fuel, and the results of sampling campaigns conducted prior to defueling indicating slight impurity gradients through the RV did not easily allow the application of a representative fuel composition to the entirety of the core during the development of the previous SFML. Further, static and accident conditions analyzed after defueling merely credited the minimum concentration of impurities to ensure the facility was safe. In each of these scenarios, the applied conservatisms are different. Currently, core debris in the lower head region of the RV is most representative of what remains in the RV at the present time. Therefore, a reasonable representative impurity concentration can be applied to the homogenized mass in development of a new SFML for D&D.Therefore, a reasonable representative impurity concentration can be applied to the homogenized mass in development of a new SFML for D&D. A conservative approach to adequately represent the inherent characteristics of the remaining fuel debris can be taken with respect to the development of an SFML for the remaining decommissioning activities. This approach would not necessarily be applicable for the previous defueling operations or the related SFML developed at that time. The current SFML was conservatively derived and, coupled with the conservatively estimated masses and the planned decommissioning operations, provides significant and adequate margin of safety that ensures that the potential for a criticality is not credible. The proposed change does not exceed or alter the SFML design basis as presented in the UFSAR and kettfor the new SFML does not exceed 0.95. Therefore, the deletion of PDMS TS 3/4.2 "Reactor Vessel Fuel" does not involve a significant reduction in a margin of safety during Phase 1band Phase 2. As part of the PDMS condition, loads in excess of 50,000 lbs. are prohibited from travel over the RV. The deletion of TS 3/4.3 does not exceed or alter a design basis or safety limit. TMl-2 Solutions will develop has developed a hoisting and rigging program that addresses movement of loads at TMl-2. The purpose of the hoisting and rigging program is to define the minimum requirements for the safe operations of cranes and hoists. The hoisting and rigging program will provide provides, as applicable, detailed requirements for training and qualification of personnel, inspection and maintenance of cranes or hoists, the safe use of rigging equipment as well as direction for performing Non-Standard Lifts in order to ensure that lifting operations are performed in a safe manner. A lift plan will be developed for all lifts as directed by the hoisting and rigging program 35 of37

Attachment 5 TMI2-RA-COR-2023-002 where a load drop or load impingement could contribute to release or dispersal of 

 *radioactive material to the environment could exceed the threshold for an unusual event.

Implementation of the hoisting and rigging program provides a defense in depth approach to preventing a load drop from occurring. Crane design features such as load cells, and travel stops, will be employed as required to ensure safe travel paths. Barriers will be provided as required to preclude the effects of a load drop. A calculation has been performed Attachment 5 (of Reference 1) in Reference 10 that assesses increasing the Safe Fuel Mass Limit (SFML) from 42 kg to approximately 1200 kg. As stated in the calculation, it is not credible to have 1200 kg U in an idealized configuration for criticality to occur. There are no credible operational upsets to realize the ideal configuration but even in the event that the upset occurs, it would require fissile mass in excess of that analyzed, which is greater than what is anticipated, in addition to a greatly reduced impurity concentration to present a criticality hazard. Therefore, the deletion of TS 3/4.3 "Crane Operations" does not significantly reduce the margin of safety during Phase 1b and Phase 2. The TMl-2 sealed sources are maintained at TMI 1 and managed by Exelon under a program compliant with the requirements of 10 CFR 70.39(c). Deleting TS 3/4.4 "Sealed Sources" from the TMl-2 TS and relocating the TS requirements to the DSAR does not involve a significant reduction in a margin of safety. The proposed changes do not affect remaining plant operations, systems, or components supporting decommissioning activities. The proposed changes do not result in a change in initial conditions, or in any other parameter affecting the course of the remaining decommissioning activity accident analysis. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. Based on the above, TMl-2 Solutions concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified. 3.4 Conclusion In conclusion, based on the considerations discussed above: 1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, 2) such activities will be conducted in compliance with the NRC's regulations, and 3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. This NSHC revision supersedes the NSHC provided in Reference 3. Change bars indicate revisions to the NSHC provided in Reference 3. Revisions made to the NSHC analysis do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c). The NSHC submitted in Reference 3 remains bounding. 36 of 37

Attachment 5 TMI2-RA-COR-2023-002

References:

1. Letter TMl2-RA-COR-2021-0002, from Van Noordennen, G. P. (TMl-2 Solutions),

License Amendment Request- Three Mile Island, Unit 2, Decommissioning Technical Specifications (ML21057A047)."

2. TMl2 Accident Analyses Questions," (ML22038A936) dated February 7, 2022
3. Letter from NRG to Sauger, J (TMl-2 Solutions, LLC), "Three Mile Island, Unit 2 -

Request for Additional Information for Requested Licensing Action Regarding Decommissioning Technical Specifications, EPID: L-2021-LLA-0038", (ML22210A087) dated July 29, 2022

4. Letter TMl2-RA-COR-2022-0007, from Van Noordennen, G. P. (TMl-2 Solutions, LLC), "License Amendment Request- Three Mile Island, Unit 2, Decommissioning Technical Specifications, Supplemental Information," (ML22101A077) dated April 8,2022
5. Letter TMl2-RA-COR-2022-0013, from Lackey, M.B. "License Amendment Request - Three Mile Island, Unit 2, Decommissioning Technical Specifications, Response to Questions", (ML22138A285) dated May 16, 2022
6. Letter TMl2-RA-COR-2022-0019, from Lackey, M.B., "License Amendment Request - Three Mile Island, Unit 2, Decommissioning Technical Specifications, Response to Request for Additional Information" (ML22276A024) dated September 29, 2022
7. NEI 99-01 Rev. 6 "Development of Emergency Action Levels for Non-Passive Reactors" dated November 2012
8. Letter 4410-90-L-0012/0477P, from GPU Nuclear, to USN RC "Defueling Completion Report, Final Submittal," dated February 22, 199" (ML20011F536).
9. Letter C312-92-2080, from GPU Nuclear, to USNRC "TMl-2 Reactor Vessel Criticality Safety Analysis," dated December 18, 1992 (ML20126D277).

10.TMl2-RA-COR-2022-0008, from Van Noordennen, G. P. (TMl-2 Solutions, LLC), 

      "Supplemental Information to License Amendment Request- Three Mile Island, Unit 2, Decommissioning Technical Specifications" (ML22108A176) dated April 7,2022
11. NRG Administrative Letter 95-06, "Relocation of Technical Specification Administrative Controls Related to Quality Assurance," dated December 12, 1995 (ADAMS Legacy Library No. 9512060318)
12. Letter TMl2-RA-COR-2022-0002, from Van Noordennen, G. P. (TMl-2 Solutions, LLC), "License Amendment Request - Three Mile Island, Unit 2, Decommissioning Technical Specifications, Supplemental Information" (ML22013A177) dated January 7, 2022
13. Masnik, M. T. (NRC) to Long, R. L. (GPU Nuclear) letter, "Issuance of Amendment No. 45 for Facility Operating License No. DPR-73 to Possession Only License for Three Mile Island Nuclear Station Unit 2 (TAC No. ML69115),"

dated September 14, 1993 (ML20029E535). 37 of37

Attachment 6 TM 12-RA-COR-2023-0002 ATTACHMENT 6 POL, TS and Bases Mark-up Pages provides a mark-up of the current POL and TS pages, including the TS Bases. This Attachment supersedes the marked-up pages provided in Reference 1. 52 Pages Follow

Attachment 6 TMI2-RA-COR-2023-002 TMI-2 SOLUTIONS, LLC DOCKET NO. 50-320 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 2 POSSESSION ONLY LICENSE Amendment No. M 67 I\ License No. DPR-73

1. The U.S. Nuclear Regulatory Commission (the NRC or the Commission) has found that:

A. The application for the transfer of the possession only license from Metropolitan \ Edison Company, Jersey Central Power and Light Company, Pennsylvania Electric Company, and GPU Nuclear, Inc. to TMI-2 Solutions, LLC (the Licensee) complies I with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code ofFederal Regulations (10 CFR) Chapter I, and all required notifications to other agencies or bodies have been duly made; B. The facility will be maintained in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission except for those exemptions from specific portions of the regulations, previously granted by the Commission, and still applicable; C. There is reasonable assurance: (i) that the activities authorized by this possession only license can be conducted without endangering the health and safety of the public; and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; D. The licensee is technically qualified to engage in the activities authorized by this possession only license in accordance with the rules and regulations of the Commission; E. The licensee is financially qualified to engage in the activities authorized by this possession only license in accordance with the rules and regulations of the Commission; F. The licensee has satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," of the Commission's regulations; G. The issuance of this possession only license will not be inimical to the common defense and security or to the health and safety of the public; Possession Only License No. DPR-73 Amendment No. M, 67

Attachment 6 TMI2-RA-COR-2023-002 H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental, and other costs and considering available alternatives, the issuance of Possession Only License No. DPR-73 subject to the conditions for protection of the environment set forth herein is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I. The possession of byproduct and special nuclear material and receipt, possession, and use of source material as authorized by the license will be in accordance with the Commission regulations in 10 CFR Parts 30, 40, and 70, including 10 CFR Sections 30.33, 40.32, 70.23, and 70.31.

2. Possession Only License No. DPR-73 is hereby issued to TMI-2 Solutions, LLC to read as follows:

A. This license applies to the Three Mile Island Nuclear Station, Unit 2 (the facility), owned by TMI-2 Solutions, LLC. The facility is located on Three Mile Island in the Susquehanna River in Londonderry Township, Dauphin County, Pennsylvania, about 10 miles southeast of Harrisburg. Prior to entry into Post-Defueling Monitored Storage (PDMS), the facility is described in the Final Safety Analysis Report as supplemented and amended, the various Recovery System Descriptions and Technical Evaluation Reports and the Environmental Report as supplemented and amended. Upon entry into PDMS, the facility is described in the PDMS Safety Analysis Report as supplemented and amended and the Environmental Report as supplemented and amended. B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses: (1) TMI-2 Solutions, LLC, pursuant to Section 103 of the Atomic Energy Act ("Act") and 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," to possess but not operate the facility; (2) TMI-2 Solutions, LLC to possess the facility at the designated location in Dauphin County, Pennsylvania, in accordance with the procedures and limitations set forth in this license; (3) TMI-2 Solutions, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any sealed sources for radiation monitoring equipment calibration; (4) TMI-2 Solutions, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) TMI-2 Solutions, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials which remain at the facility subsequent to the cleanup following the March 28, 1979, accident. Possession Only License No. DPR-73 Amendment No. 64, 67

Attachment 6 TMI2-RA-COR-2023-002 The storage ofradioactive materials or radwaste generated at TMI Unit 1 and stored at TMI Unit 2 in accordance with the license for TMI Unit 1 shall not result in a source term that, if released, would exceed that previously analyzed in the PDMS Safety Analysis Report in terms of off-site dose consequences. C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I, and is subject to all applicable provisions of the Act and to the Commission's rules and regulations, except for those exemptions from specific portions of the regulations granted by the Commission and still applicable, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Technical Specifications The Technical Specifications, as revised through Amendment No. ee67 are hereby incorporated into this license. The licensee shall operate maintain the facility in accordance with the Technical Specifications and all Commission Orders issued subsequent to the date of the possession only license. (2) Physical Protection TMI-2 utilizes a security plan (TMI-2 Materials Security Plan) that is compliant with 10 CFR Part 3 7, "Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material." The plan will implement 10 CFR Part 3 7 security requirements and also implement specific 10 CFR 73 .67 security requirements referenced in the security plan that are not addressed by Part 37. (3) Upon the date of closing, and proceeding until determination of completion of Phase 2 of facility decommissioning, TMI-2 Solutions will maintain a Financial Support Agreement in the amount of $1 OOM, less the value of any cash-funded Provisional Trust Account, Disposal Capacity Easement, and Letter of Credit procured by TMI-2 Solutions for the benefit of the Back-Up Trust Account under the Back-Up & Provisional Trust Agreement. (4) At time of closing, EnergySolutions, Inc. will provide a Parent Guarantee in favor of the FirstEnergy Companies to guarantee the payment and performance of the obligations of TMI-2 Solutions as to the TMI-2 decommissioning. This guarantee makes the resources of EnergySolutions available to help ensure the successful decommissioning of TMI-2, assuring the ability of TMI-2 Solutions to (i) pay the costs of decommissioning the TMI-2 facility; (ii) protect the public health and safety; and (iii) meet NRC requirements. (5) These financial support conditions (2.C.(3) and 2.C.(4)) may not be voided, canceled, or modified without the prior written consent of the NRC. These financial support conditions are in place and will be maintained as described in the application. The Director of the Office of Nuclear Material Safety and Safeguards shall be informed, in writing, no later than 10 working days after any funds are provided under the terms of the conditions listed above. Possession Only License No. DPR-73 AmendmentNo.64,e$-,67

Attachment 6 TMI2-RA-COR-2023-002 (6) TMI-2 Solutions, the licensee for TMI-2, will not perform major decommissioning activities which would diminish the historic integrity of the TMI-2 owned and controlled buildings until the applicable historic and cultural reviews are completed by the Nuclear Regulatory Commission. D. Special Am1.iliary and Fuel Handling Building Ventilation Study: Prior to terminating continuous operation of the amdliary and fuel handling buildings (AFHB) ventilation systems, the special monitoring program ofAFHB airborne levels shall be completed. The program shall include at least 1 year of data prior to entry into PD~4S and at least 1 year of data after entry into PD~4S. A report shall be submitted to the 1'tR.C containing the results of the program eontaining sufficient data and analyses to demonstrate that the release rate of particulates v<'ith half lives greater than 8 days from the AFHB 1Nill be less than 0.00024 µCi/sec when averaged over any calendar quarter. Not included in the calculation of the particulate release rate shall be those periods oftime (designated in advance) prior to entry into PDMS during ,vhich aggressive decontamination operations were performed in preparation for PD~4S. The report shall be submitted to the NRG staff at least 60 days prior to terminating continuous operation of the AFHB ventilation systems. E. Unfiltered Leak Rate Test: Prior to entry of the facility into Post Defucling Monitored Storage, the licensee v,ill develop an NRG approved surveillanee requirement for the reactor building IB1filtered leak rate test that, upon staff approval, will be incorporated as Section 4.1.1.2 of the proposed PDMS Technical Specifications. F. Additional Submittals Prior To Post Defueling Monitored Storage: Prior to entry of the facility into Post Defueling Monitored Storage, the licensee will submit and implement a site Flood Protection Plan, a site Radiation Protection Plan, an Offsite Dose Calculation Manual, a Post Defueling Monitored Storage Fire Protection Program Evaluation, a Post Defueling Monitored Storage Quality A.ssurance Plan and a Radiological Environmental Monitoring Plan. Additionally, the licensee will submit to the NRG the results of the completed plant radiation and contamination surveys prior to entry into PDMS. G. This license is effective as of the date of issuance and until the Commission notifies the licensee in writing that the license is terminated. FOR THE NUCLEAR REGULATORY COMivJISSION (Original signed by Alfred E. Chaffee acting for) Brian K. Grimes, Director Division of Operating Reactor Support Office of Nuclear Reactor Regulation

Enclosure:

Appendixses-A &-B Technical Specifications Date of Issuance: December 18, 2020 Possession Only License No. DPR-73 AmendmentNo.64,{8., 67

Attachment 6 TMI2-RA-COR-2023-002 SECTION 1.0 DEFINITIONS

Attachment 6 TMI2-RA-COR-2023-002 1.0 DEFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications. POST DEFUELil'JG MONITORED STORAGE 1.2 POST DEFUELil'JG MONITORED STORAGE (PDMS) is that condition where TMI 2 defueling has been completed, the core debris removed from the reactor during the cleanup period has been shipped off site and the facility has been placed in a stable, safe, and secure condition. A.CTION 1.3 ACTION shall be those additional requirements specified as corollary statements to each specification and shall be part of the specifications. OPERABLE OPERABILITY

1. 4 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s) and when all necessary attendant instrumentation, controls, electrical pov1er, cooling or seal 1.vater, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related SUfJpOrt function(s).

CHMJNEL CALIBRATION 1.5 An instrument CHAN1'l"EL CALIBRATION is a test, and adjustment, as necessary, to establish that the ehannel output responds with acceptable rang and accuracy to lrnown values of the parameter 1.vhich the channel measures or an accurate simulation of these values. CH,<\l'JNEL CALIBRATION shall encompass the entire channel including equipment activation, alarm or trip, and shall be deemed to include the CHANNEL FillJCTIONAL TEST. CH,<\l'JNEL CHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status vlith other indications and/or status derived from independent instrument channels measuring the same parameter. CHMJNEL FillJCTIONAL TEST

1. 7 CHi<\l'JNEL FillJCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERA,BILITY including alarm and/or trip functions.

Three Mile Island - Unit 2 1-1 Amendment48-,67

Attachment 6 TMI2-RA-COR-2023-002 1.0 DEFINITIONS FREQUENCY NOTATION 1.8 The FREQUfil,JCY NOTi\Tirn-J specified for the perfonnance of surveillance requirements shall correspond to the intervals defined in Table 1.1. CONTA,Il'{MfilJT ISOL'\TION 1.9 CONTAINMENT ISOLl\TION shall eJtistwhen:

a. Each penetration is:
1. Closed by a manual valve, a .velded or bolted blind flange, a deactivated automatic 1

valve secured in a closed position or other equivalent mechanical closure to provide isolation of each penetration, or

2. Open and the pathway to the environment provided with HEPA filter, or
3. Open in accordance 1,vith approved procedures. Controls shall be implemented to minimize the time the penetration is allov1ed open and to specify the conditions for which the penetration is open. Penetrations shall be rnq3editiously closed upon completion of the conditions specified in the approved procedures, and
a. The Equipment Hatch is closed, and
b. Each Containment Airlock is operable pursuant to Technical Specification 3 .1.1.3.

BfJCH RELEASE 1.10 A BATCH RELEl'.cSE is the discharge of a discrete volume. CONTIJ\JUOUS RELEASE 1.11 A crn-JTil'JUOUS RELEASE is the discharge of a non discrete volume, e.g., from a volume or system that has an input flow during the continuousrelease. OFF-SITE DOSE CALCULATION MANUAL 1.12 OFF-SITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of off-site doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the programs required by Section 6.7.4 the Decommissioning Quality Assurance Plan and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating* and Annual Radioactive Effluent Release Reports required by Specifications 6.8.1.2 the Decommissioning Quality Assurance Plan and Specification 6.8.1.2, respectively. Three Mile Island - Unit 2 1-2 Amendment48-,67

Attachment 6 TMI2-RA-COR-2023-002 1.0 DEFINITIONS REPORTABLE EVENTS 1.13 ,\ REPORL'\BLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Pa-rt 50. SLA..GGERED TEST BASIS 1.11 A STAGGERED TEST BA.SIS shall consist of:

a. A test schedule for n systems, subsystems, trains or designated components obtained by dividing the specified test interval into n equal subintervals,
b. The testing of one system, subsystem, train or designated components at the beginning of each subinterval.

SUBSTANTI\'E CHANGES 1.15 SUBSTANTI\'E CHAJ\l"GES are those 1.vhich affect the activities associated vi1ith a document or the document's meaning or intent. Examples of non substantive changes are: (1) correcting spelling; (2) adding (but not deleting) sign off spaces; (3) blocking in notes, cautions, etc.; (4) changes in corporate and personnel titles which do not reassign responsibilities and v.1rich are not referenced in the PDMS Technical Specifications; and (5) changes in nomenclature or editorial changes v1hich clearly do not change function, meaning or intent. Three Mile Island- Unit 2 1-3 Amendment 48-, 67

Attachment 6 TMI2-RA-COR-2023-002 1.0 DEFINITIONS l\4EMBER(8) OF THE PUBLIC 1.16 }.4E:l\1BER(s) OF THE PUBLIC means any individual except v.r.hen that individual is receiving an occupational dose. UNRESTRICTED AREA 1.17 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUJ\JDARY access to ,,:vhich is not controlled by TMI 2 Solutions, LLC for pm-poses of protection of Individuals from exposure to radiation and radioactive materials, or any area within the SITE BOillJDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. SITE BOUNDARY 1.18 The SITE BOillJDA.RY shall be that line beyond ,vhich the land is neither owned, nor leased, nor otherwise controlled by TMI 2 Solutions, LLC. The SITE BOillJDA.RY for gaseous and liquid effluents shall be as shovm in the Offsite Dose Calculation Manual (ODCM). N::PDES PER}.4IT 1.19 The J'WDES PER.1\4IT is the National Pollutant Discharge Elimination System (]'WDES) Permit No. PA0009920, effective January 30, 1975, Issued by the Environmental Protection Agency to }.4etropolitan Edison Company. This permit authorized }.fotropolitan Edison Company to discharge controlled ,.vastev,rater from Three Mile Island (T}.41) Nuclear Station Into the 1Naters of the Commonwealth of Pennsylvania. Three Mile Island- Unit 2 1-4 Amendment 48-, 60, ~ 67

Attachment 6 TMI2-RA-COR-2023-002 TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY

g. At least once per 12 hours.
g At least once per 24 hours.

w At least once per 7 days. M At least once per 31 days. Q At least once per 92 days. SA At least once per 184 days. A At least once per 12 months R At least once per 18 months

p. Completed prior to each release.

NIA Not applicable. Three Mile Island - Unit 2 1-5 Amendment 48-, 67

Attachment 6 TMI2-RA-COR-2023-002 SECTION 2.0 SAFETY LIMITS

Attachment 6 TMI2-RA-COR-2023-002 2.0 SAFETY LIMITS There are no safety limits which apply to TMI 2 during PDMS. Three Mile Island - Unit 2 2-1 Amendment 48, 67

Attachment 6 TMI2-RA-COR-2023-002 SECTION 3/4 Lil\fETING CONDITIONS FOR PDl\4S SURVEILL1A~CE REQUIREl\4ENTS

Attachment 6 TMI2-RA-COR-2023-002 3/4 .0 LIMITING CONDITIONS FOR PDMS ,'\J'JD SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LilvUTil'tG CONDITIONS FORPDMS 3.0.1 Limiting Conditions for PDMS and ACTION requirements shall be applicable during POST DEFUELING MONITORED STORAGE or other conditions specified for each specification. 3 .0.2 Adherence to the requirements of the Limiting Condition for PDMS and/or associated ACTION 1.vithin the specified time interval shall constitute compliance with the specification. In the event the Limiting Condition for PDMS is restored prior to e1rpiration of the specified time interval, completion of the ACTION statement is not required. 3.0.3 In the event a Limiting Condition for PDMS and/or associated i\CTI01'l" requirements cannot be satisfied because of circumstances in e1ccess of those addressed in the specification, initiate appropriate actions to rectify the problem to the extent possible under the circumstances and submit a repmi to the Commission pursuant to the requirements of 10 CFRS0.73. SURVEILL'\J'JCE REQUIREMENTS 4 .0.1 Surveillance Requirements shall be met during PDMS or other conditions specified for individual Limiting Conditions for PDMS unless otherwise stated in an individual Surveillance Requirement. 4.0.2 Each Surveillance Requirement shall be performed 1.vithin the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval. 4.0.3 Failure to perform a Surveillance Requirement v:ithin the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for PDMS. Exceptions to these requirements are stated in the individual Specifications. Surveillance Requirements do not have to be performed on inoperable equipment. 4.0.4 Ifit is dissevered that a surveillance was not performed .vithin its specified frequency, then 1 compliance 1.vith the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours orup to the limit of the specified frequency, 1.vhichever is less. This delay period is permitted to allow performance of the Surveillance.

  • I' Three Mile Island - Unit 2 3/4.0-1 Amendment 1-, 67

Attachment 6 TMI2-RA-COR-2023-002 3/4 .1 CONTA.IN1ffiNT SYSTEMS 3/4 .1.1 PRIMA.RY CONTAJNMENT CONT.t\INMENT ISOLATION LIMITING CONDITIONS FORPDMS 3 .1.1.1 Primary CONTJ\IN1ffiNT ISOLATION shall be maintained. APPLICABILITY: PDMS ACTiilll": With CONTAINMENT ISOLATiill-J not in aeeordanee with requirements, restore CONT1...IN1ffiNT ISOLATION ,vithin 24 hours. SURVEILLANCE REOUIRE1ffiNTS 4.1.1.1 Primaiy COOTAINMEJ'JT ISOLATiill-J shall be verified quarterly with the following e:x:esptions:

a. Isolation valves that arc locked closed shall be verified annually on a quarterly STAGGERED TEST BASIS. If a valve is found to be out of position, a eheek of all looked dosed isolation valves shall be performed.
b. An independent verification of all isolation valve position changes shall be performed.
e. Bolted or welded blind flanges which form a containment isolation boundary and the Equipment Hateh shall be visually inspeeted for signs of degradation and/or leakage every five years on an annual SL<\.GGERED TEST BASIS. If a problem is dissevered with a flange, a ehcck of all bolted or ,velded blind flanges shall be performed.

Three Mile Island - Unit 2 3/4.1-1 Amendment 48, 67

Attachment 6 TMI2-RA-COR-2023-002 UNFILTERED LEAK RATE TESTING LIMITING CONDITIONS FORPDMS 3 .1.1.2 The unfiltered leak rate from Containment .vith the RB Breather closed shall be less than 1 1/100 of the rate ilirough the RB Breather. APPLICABILITY: PDMS ACTION: If the unfiltered leak rate from Containment with the RB Breather closed is greater than 1/100 of the rate tlwough the RB Breather or if the trend indicates that the 1/100 value 1.vill be exceeded 1.vithin one year, then:

a. Identify the eJwessive leakagepath;
b. Make necessary repairs and/oradjustments; C. Perform an additional unfiltered leak rate test; and
6. Prepare and submit a special report to the Commission pursuant to Specification 6.8.2 within the neJct 30 days.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The initial unfiltered leak rate test shall be performed r.vo years following entry into PDMS. After the initial unfiltered leak rate test, the test frequency will be determined by comparing the ratios of the unfiltered leak rate to the RB Breather leak rate from previous and current tests. If the test results indicate that the ratio of unfiltered leakage to breather leakage is remaining constant or decreasing, then the next interval shall be five years. Three Mile Island - Unit 2 3/4.1-2 Amendment 48, 67

Attachment 6 TMI2-RA-COR-2023-002 SURVEILLANCE REQUIREMENTS 4.1.1.2 (con't) If the test results indicate that the ratio of unfiltered leakage to breather leakage is increasing, i.e., the current ratio is greater than the previous ratio, then the next interval shall be determined by the follo'tving equation: 0.01 whern:N' the next test interval, 

       }l the current test interval,

--Ip the previous ratio of unfiltered leakage to RB Breather leakage - - R the current ratio of unfiltered leakage to RB Breather leakage The initial value ofN shall equal two years. N' shall be the truncated integer result from the above equation, in years, but not more than five years nor less than one year. Only ratios for successful tests shall be used to determine the next test interval in the above equation. Following a failed test the ne~ct test interval shall be one year. Three Mile Island - Unit 2 3/4.1-3 Amendment 48, 67

Attachment 6 TMI2-RA-COR-2023-002 CONTAINMENT AIR LOCKS SURVEILLANCE REQUIREMENTS 3 .1.1.3 Each Contaimncnt Air Lock shall be OPERA::BLE v,:ith at least one door closed except when the air lock is being used for transit entry and exit in accordance 1 

.vith site approved procedures.

APPLICABILITY: PDMS ACTION: Vlith no Containment Air Lock door OPERABLE, restore at least one door to OPERABLE status within 24 hours. SURVEILLAJ'JCE REOUIRE+/-\ffiNTS 4.1.1.3 Each Containment Air Lock shall be demonstrated OPERABLE every five (5) years by performing a mechanical operability check of each Air Lock Door, including a Yisual inspection of the components and lubrication if necessary and by visually inspecting the door seals for significant degradation. \\Then both Containment Air Lock doors are opened simultaneously, verify the following conditions:

a. The capability exists to m,peditiously close at least one Air Lock door;
b. The Air Lock doors and Containment Purge are configured to restrict the outflov,r of air in accordance .vith site approved procedures; and 1

C. The Air Lock doors are cycled to ensure mechanical operability v1ithin seven days prior to opening both doors. Three Mile Island - Unit 2 3/4.1-4 Amendment 51, 61, 67

Attachment 6 TMI2-RA-COR-2023-002 3/4 .2 REf.cCTOR VESSEL FUEL 3/1.2.1 REl...CTOR VESSEL FUEL REMOVAL/REA.RR,".tNGEMENT LIMITING CONDITIONS FORPDMS 3.2.1.1 No more than 12 kg of fuel (i.e., U02) maybe removed from the Reactor Vessel without prior NRG approval. APPLICABILITY: PDMS AcCTION: Vlhen more than 12 kg of fuel has been removed from the Reactor Vessel, suspend all further fuel removal activities and submit a safety analysis to the NR{; for approval of this activity and any further fuel removal activities. 3.2.1.2 No more than 42 kg of fuel in the Reactor Vessel may be rearranged outside the geometry's analyzed in the Defueling Completion Report and the criticality safety analyses contained in GPU Nuclear letter C312 92 2080, dated December 18, 1992, without prior NR{; approval. APPLICABILITY: PDMS ACTION: '}lhon more than 42 kg of fuel in the Reactor Vessel has been rearranged, suspend all further fuel rearrangement activities and submit a safety analysis to the NRG for approval ofthis activity and any further fuel rearrangement activities. ffan external event were to occur that could potentially cause more than 12 kg of fuel in the Reactor Vessel to be rearranged, a report 1.vill be submitted to the }'JR{; detailing the findings of any investigation into that potential rearrangement. SURVErLLA1'l"CE REQUIREMENTS 4.2.1.1 None required as long as no fuel is removed from the ReactorVessel. 1.2.1.2 None required as long as no fuel in the Reactor Vessel is rearranged. Three Mile Island - Unit 2 3/4.2-1 Amendment 48, 67

Attachment 6 TMI2-RA-COR-2023-002 3/4 .3 CRANE OPERATIONS Ll1\4ITING CONDITIONS FORPDMS 3.3.1 Loads in excess of 50,000 lbs. shall be prohibited from travel over the Reactor Vessel unless a docketed Safety Evaluation for the activity is approved by the NRG. APPLICABILITY: PDMS i\CTION: 'Nith the requirements of the above specification not satisfied, place the crane load in a safe condition and conect the circumstances which caused or allmved the Limiting Condition for PDMS to be exceeded prior to continuing crane operations limited by Specification 3 .3 .1. Prepare and submit a special report to the Commission pursuant to Specification 6.8.2 1,vithin the nffict 30 day&; Three Mile Island - Unit 2 3/4.3-1 Amendment 48, 67

Attachment 6 TMI2-RA-COR-2023-002 3/4.4 SEALED SOUR{:;ES 3/4 .4 .1 SEALED SOUR{:;E INTEGRITY LIMITING CONDITIONS FOR PDMS

3. 4.1 Each sealed source containing radioactive material either in e1wess of 2100 micro curies of beta and1or gamma emitting material or 5 microcuries of alpha emitting material (except as noted in 4.4 .1.2) shall be free oL::::0.005 micro curies of removable contamination.

APPLICABLE: PDMS ,\CTION:

a. Each sealed source 1.vith removable contamination in mwess of the above limit shall be immediately 1.vithdrawn from use and:
1. Either decontaminate and repair, or
2. Dispose in accordance with CommissionRegulations.
b. The provisions of Specification 3.0.3 are not applicable.

SURVEILL'\l'tCE REQUIREMENTS TEST REOUIRElVIENTS 4.4.1.1 Each sealed source shall be tested for leakage and1or contamination by:

a. The licensee, or
b. Other persons specifically authorized by the Commission or an A,greement State.

The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample. TEST FREQUENCIES 4.4.1.2 Each category of sealed source shall be tested at the frequency described belmv. a Source in use (excluding fission detectors previously subjected to core flux) At least once per sh( months for all sealed sources containing radioactive material:

1. ',¥ith half life greater than 30 days (excluding Hydrogen 3) and
2. In any form other than gas.

Three Mile Island - Unit 2 3/4.4-1 Amendment 48, 67

Attachment 6 TMI2-RA-COR-2023-002 SURVEILLAl'tCE REQUIREMENTS

b. Stornd soUrces not in use Eaoh sealed souroe and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous six months. Scaled sources and fission detectors transferred 1.vithout a certificate indicating the last test date shall be tested prior to being placed into use.

C. Fission detectors Each sealed fission detector shall be tested 1.vithin 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source. REPORTS 4.4 .1.3 A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of::::0.005 microcuries ofremovable contamination. Three Mile Island - Unit 2 3/4.4-2 Amendment 48, 67

Attachment 6 TMI2-RA-COR-2023-002 BASES F-QR LThflTIN:O CONDITIONS FOR PDMS MID SURVEILLANCE REQUIRElvffiNTS

Attachment 6 TMI2-RA-COR-2023-002 NOTE The summrn=y statements contained in this section provide the bases for the Specifications of Section 3 .0 and 4.0 and are not considered a part of these Technical Specifications as provided in 10 CPR 50.36.

Attachment 6 TMI2-RA-COR-2023-002 3/4.0 APPLICABILITY The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for PDMS and Survoillanoe Requirements within Seotion 3/4. 3 .0.1 This speoifioation defines the applioability of eaoh ~peoifioation in terms of PDMS or other speoified oonditions and is provided to delineate speoifioally 1vvhen eaoh speoification is applioable. 3.0.2 This speoifioation defines those oonditions neoessary to oonstitute oomplianoe with the terms of an individual Limiting Condition for PDMS and assooiated ACTION requirement.

3. 03 This specification defines the action and reporting requirements for those oircumstanoes vmere the ACTION statement for Limiting Conditions for PDMS 1.vas exoeeded.
4. 0 .1 This specifioation provides that the surveillanoe aotivities necessary to ensure the Limiting Conditions for PDMS am met and will be performed during the oondition for which the Limiting Conditions for PDMS are applioable.

4.0.2 The provisions of this specification provide allowable toleranoes for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances am necessary to provide operational flmdbility because of scheduling and performance considerations. The phrase "at least" associated vlith a surveillance frequency does not negate this allowable tolerance value and permits the performance ofmore frequent surveillance activities. It is not intended that this provision be used repeatedly as a convenience to mctend surveillance intervals beyond that specified. The allowable toleranoe is based on engineering judgement and the recognition that the most probable result of any partioular surveillance being performed is the verifioation of oonformanoe 1.vith the Surveillanoe Requirements. This provision is suffioient to ensure that the reliability ensured through surveillanoe activities is not significantly degraded beyond that obtained from the speoified surveillanoe interval. 4.0.3 The provisions of this speoifioation set forth the oriteria for determination of oompliance with the OPERABILITY requirements of the Limiting Conditions for PDMS. Under this oriteria, equipment, systems or oomponents are assumed to be OPERABLE if the associated surveillanoe aotivities have been satisfaotorily performed 1.vithin the speoified time interval. Nothing in this provision is to be oonstrued as defining equipment, systems or components as OPERABLE, when suoh items are found or known to be inoperable although still meeting the Surveillance Requirements Three Mile Island - Unit 2 B3/4.0-1 Amendment -M, 67

Attachment 6 TMI2-RA-COR-2023-002 3/4 .0 APPLICABILITY (Con't)

4. 04 This specification establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a surveillance has not been completed 1.vithin the specified frequency. A delay period of up to 24 hours applies from the point in time that it is discovered that the required surveillance has not been performed and not at the time that the specified frequency was not met.

The delay period provides an adequate time to complete sur,reillances that have been missed. This delay period ponnits the completion of a surveillance before complying with required actions or other remedial measures that might preclude completion of the surveillance. The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the surveillance, the safety significance of the delay in completing the required surveillance, and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the requirements. 'Nhen a surveillance *.vith a frequency based not on time intervals, but upon specified unit conditions or operational situations, is discovered not to have been performed when specified, this provision allows the full delay period of 24 hours to perform the surveillance. Failure to comply 1.vith specified surveillance frequencies is expected to be an infrequent occurrence. Use of the delay period is not intended to be used as an operational convenience to eJctend surveillance intervals. If a surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the completion times of the required actions for the applicable LCO conditions begin immediately upon expiration of the delay period. If a surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the completion times of the required actions for the applicable LCO conditions begin immediately upon failure of the surveillance. Completion of the surveillance with the delay period all01.ved by this specification, or within the completion time of the actions, restores compliance. Three Mile Island - Unit 2 B3/4.0-2 Amendment*, 67

Attachment 6 TMI2-RA-COR-2023-002 3/4 .1 CONTAINMENT SYSTEMS 3/4 .1.1 PRIMARY CONTAINMENT 3/4 .1.1.1 CONTAIN~@JT ISOLATION CONTA,IN-~@JT ISOLATION is maintained to assure the Containment is properly maintained as a contamination barrier for the residual contamination which remains inside the Containment. One barrier either outside or inside of the Containment on each penetration is acceptable. See the PDMS SAR Section 7.2.1.1. Verification of CillJLt\Il'JMENT ISOLATiillJ is primarily accomplished by visual inspection; hov;ever, in cases where this is not practical due to the valve or valves being located in a locked high radiation area, documented e11idence of the valves closure may be used. Penetrations which have been isolated by chain locked valves provide a high degree of assurance that CONTAIN~@JT ISOLATION is being maintained and, therefore, require only annual surveillance on a STAGGERED TEST BASIS. Penetrations 1.vhich have been closed by bolted or 1.velded blind flanges provide an even higher degree of assurance that CONL'\Il'JMENT ISOLATION is being maintained and, therefore, require surveillance only every five years also on a STAGGERED TEST BASIS. However, if a valve is found out of position or a problem .vith a 1 flange is discovered, a complete verification check 1.vould be performed to provide assurance that CONL'\Il'JMENT ISOLA.TiillJ is being maintained. 3/4 .1.1.2 UJ'ffILTERED LEAK Rl,:TE TESTil'JG The R-0aetor Building fire analysis presented in SAR Section 8.2.5 Case 3 assumes that the mass flmvrate of unfiltered leakage is less than 1/100 of the mass flowrate released through the 99% efficient RB Breather HEPA filter. SAR Section 7.2.1.2.3 provides the details of the calculation using an unfiltered leak rate test to demonstrate compliance with this Limiting Condition for PDMS. The test interval is variable due to the uncertainty inherent in maintaining the unfiltered leakage to a small fraction of the leakage through the RB Breather. 3/4 .1.1.3 CONTA.Il'J~@JT AIRLOCKS The Containment Air Locks must be maintained OPERABLE to provide CONTA.Il'JMENT ISOLATiillJ. These air locks v:ill be used during entries into the Containment to ensure that radioactive materials are not unnecessarily being released to the environs. The preferred method for ensuring that radioactive materials are not released during these entries is to maintain at least one door closed at all times; however, if circumstances require, both doors may be open simultaneously in accordance with site approved procedures. Three Mile Island - Unit 2 B3/4.1-1 Amendment 48, 67

Attachment 6 TMI2-RA-COR-2023-002 3/4.2 REACTORVESSELFUEL 3/4.2.1 REACTOR VESSEL FUEL REMOVAL/REARRA.NGE~ffiNT NRG Inspection Report 50 320/90 30, dated June 14, 1990, imposed restrictions on the removal and/or rearrangement of the residual fuel in the Reactor Vessel. In particular, the NR{? stated in Section 3 .0, "Safe Fuel Mass Limit," of that inspection report that the appropriate safe fuel mass limit in the Reactor Vessel (RV) was determined to be 93 kg of core debris. Based on industry practice, a limit of approximately 45% of the SFML was placed on the amount of core debris that may be removed from the RV or rem.Tanged in the RV. This limit is specified to ensure subcriticality even after dual errors. Thus, if the fuel in the RV is rearranged outside the analyzed geometries used in the Defueling Completion Report or the criticality safety analyses contained in GPU J\Juclear letter C312 92 2080, dated December 18, 1992, the 42 kg limit will apply to the rearranged fuel. Further, if any fuel is removed from the RV in the future, the 42 kg limit will also apply to that fuel. Three Mile Island - Unit 2 B3/4.4-1 Amendment 48, 67

Attachment 6 TMI2-RA-COR-2023-002 3/4 .3 CRANE OPERATIONS A load drop into the RV may cause reconfiguration of the eore debris outside the analyzed geometries used in the Defueling Completion Rep mi RV criticality analysis. Three Mile Island - Unit 2 B3/4.4-1 Amendment 48, 67

Attachment 6 TMI2-RA-COR-2023-002 3/4. 4 SEALED SOUR{;ES 3/4 .4 .1 SEALED SOURCE INTEGRITY The limitation on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation ,,vill ensure that leakage from byproduct, source, and Special }'fuclear Material sources 'Nill not mweed allov,rable intake values. Three Mile Island - Unit 2 B3/4.4-1 Amendment 4&, 67

Attachment 6 TMI2-RA-COR-2023-002 SECTION 5.0 DESI@J FEATURES

Attachment 6 TMI2-RA-COR-2023-002 5.0 DESIGN FEATURES 5.1 CONTAIN1\4ENT CONFIGURi\TION 5 .1.1 The Containment Building is a steel lined, reinforced concrete building of cylindrical shape, 1.vith a dome roof and having the foll01.ving design features:

a. Nominal inside diameter 13 0 feet.
b. Nominal inside height 157 feet.

C. Minimum thickness of concrete walls 4 feet.

d. Minimum thickness of concrete roof- 3. 5 feet.
e. Minimum thickness of concrete floor pad 13. 5 feet.

£ Nominal thickness of steel liner - 1/2 inches.

g. Net free volume 2.1 Jc. 106-cubic feet.
h. Design Pressure 5.0 psig.

Three Mile Island - Unit 2 5-1 Amendment 48, 67

Attachment 6 TMI2-RA-COR-2023-002 SECTION 6.0 ADMINISTRATIVE CONTROLS Three Mile Island - Unit 2 6-1 Amendment48,54,59,67

Attachment 6 TMI2-RA-COR-2023-002 6.0 ADMINISTRATIVE CONTROLS

u. Responsibility 6.1.l The Tl\11 2 Solutions, LLC Project Director is responsible for the management of overall unit operations at Unit 2 and shall delegate in writing the succession to this responsibility during absence.

ORDANIZATION TMI 2 SOLUTIONS ORDA}lIZl.,TION 6.2.1 The Tl\11 2 Solutions, LLC organization for unit management and technical support shall be as in Section 10.5 of the PD1\1S SAR. TMJ 2 SOLUTIONS UNIT ORGANIZATION 6.2.2 The unit organization shall be as described in Section 10.5 of the PDMS SAR and an individual qualified in radiation protection procedures shall be on site whene:ver Radioactive \Vaste Management activities are in progress. 6.3 UNIT STAFF OUALIFICi\TlillJS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of A..NSI NI 8.1 1971 for comparable positions unless otherwise noted in the Technical Specifications. The requirements of ,'\NSI N18.l 197lthat pertain to operator license qualifications for unit staff shall not apply. 6.3.2 The management position responsible for radiological control or his deputy shall meet or exceed the qualifications of Regulatory Guide 1.8 of 1977. Each Radiological Controls Technician in a responsible position shall meet or e~rneed the qualifications of ANSI N18.l 1971, paragraph

4. 5 .2 or 4 .3 .2, or be formally qualified through an NRG approved Tl\11 Radiation Controls training program. All Radiological Controls Technicians will be qualified through training and examination in each area or specific task related to their radiological controls functions prior to their performance of those tasks.

6.4 DELETED Three Mile Island- Unit 2 6-2 Amendment48-,M,-5-9-,67

Attachment 6 TMI2-RA-COR-2023-002 ADMIN]STRATIVE CilllTROL8 6.5 DELETED 6.5.1 DELETED DELETED Three Mile Island- Unit 2 6-2 Amendment 48-, -54, 6J, 67

Attachment 6 TMI2-RA-COR-2023-002 ADMil>HSTRfJIVE CONTROLS DELETED DELETED 6.5.2 DELETED DELETED Three Mile Island- Unit 2 6-3 Amendment 0, M, fH., ~ 67

Attachment 6 TMI2-RA-COR-2023-002 ADMINISTRATIVE CONTROLS DELETED Three Mile Island- Unit 2 6-4 Amendment 48-, .§4, ~ . ~ 67

Attachment 6 TMI2-RA-COR-2023-002 ADMINISTRATIVE CONTROLS DELETED DELETED 6.5.3 DELETED Three Mile Island- Unit 2 6-5 Amendment 2-, ~ 67

Attachment 6 TMI2-RA-COR-2023-002 ADMINI8TRA.TIVE CONTROLS DELETED 6.5.4DELETED Three Mile Island- Unit 2 6-6 Amendment-5-2-,M,~,@,67

Attachment 6 TMU-RA-COR-2023-002 ADMINISTRATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The 1'ruclear Regulatory Commission shall be notified and/or a report submitted pursuant to the requirements of Section 50.73 to 10 CFR 50.

6.7 PROCEDURES AJ\ID PROGRAMS 6.7.1 V/ritten procedures shall be established, implemented, and maintained for the activities necessary to maintain the PDMS condition as described in the PDl\1S SAR. Examples of these activities are:

a. Technical Specification implementation.
b. Radioactive *.vaste management and shipment.

C. Radiation Protection Plan Implementation.

d. Fire Protection Program implementation.
e. Flood Protection Program implementation.

Three Mile Island - Unit 2 6-7 Amendment 48, 54,59,@,67

Attachment 6 TMI2-RA-COR-2023-002 ADlMINISTRATIVE CONTROLS 6.7 PROCEDURES A1'ID PROGRP.cMS (cont'd) 6.7.2 Each procedure required by Section 6.7.1, and SUBSTAJ,JTIVE CW..NGES thereto, shall be reviewed and approved prior to implementation and shall be reviev,,ed periodically as follows:

a. .At least every two years, the group responsible for Quality Assurance ,vill assess a representative sample of plant procedures that are used more frequently than every two years.
b. Plant procedures that have been used at least biennially receive scrutiny by individuals knowledgeable in procedures, and are updated as necessary to ensure adequacy during suitable controlled activities.
e. Plant procedures that have not been used for tvw years will be reYievred before use or biennially to determine if changes are necessary or desirable.

6.7.3 Temporary changes to procedures in Section 6.7.1 above may be made provided:

a. The intent of the original procedures is not altered;
b. The change is approved by two members of the responsible organization kno*,vledgeable in the anm affected by the procedure. For changes which may affect the operational status of unit systems or equipment, at least one of these individuals shall be a member of unit management or supervision; and
e. The change is documented, revievred and approved within 14 days of implementation.

6.7.4 The follo*,ving programs shall be established, implemented, and maintained:

a. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to 1\ffiMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCl\4, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded.

The program shall include the following elements: Three Mile Island - Unit 2 6-8 Amendment 48, 56, 63, 67

Attachment 6 TMI2-RA-COR-2023-002 ADMINISTRATIVE CONTRDLS 6.7PROCEDURES l'.c-ND PROGRAMB (cont'd)

1. Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,
2. Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED A.REAS conforming to 10 times the concentrations specified in 10 CFR Part 20.1001 20.2402, Appendbc B, Table 2, Column 2,
3. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance 1.vith 10 CFR 20.1302 and vlith the methodology and parameters in the ODCM, Three Mile Island- Unit 2 6-8a Amendment 56, 60,67

Attachment 6 TMI2-RA-COR-2023-002 ADMIJ\HSTRATIVE CONTROLS 6.7 PROCEDURES AND PROGRAJVIS (cont'd)

4. Limitations on the annual and quarterly doses or dose commitment to a l\ffil\ffiER OF THE PUBLIC from radioactive materials in liquid effluents released from the unit to the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,
5. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance \Vith the methodology and parameters in the ODCM at least evef)' 31 days,
6. Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity vvhen the projected doses in a 31 day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,
7. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas at or beyond the SITE BOUNDARY. The limits are as follows:

a) For noble gases: less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin, and b) For tritium and all radionuclides in particulate form *.vith half lives greater than 8 days: less than or equal to 1500 mrem/yr to any organ,

8. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from the unit to areas beyond the SITE BOillIDARY conforming to Appendbc I to 10 CFR Part 50,
9. Limitations on the annual and quarterly doses to a l\ffil\ffiER OF THE PUBLIC from tritium and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOillIDARY conforming to Appendix I to 10 CFR Part 50,
10. Limitations on the annual dose or dose commitment to any l\ffil\ffiER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle soUrces conforming to 40 CFR Part 190.
a. Radiological Emrironmental Monitoring Program A program will be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representatiYe measurements of radioactivity In the highest potential e)cposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposUre pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) Include the following:

Three Mile Island- Unit 2 6-9 Amendment 40, 60,67

Attachment 6 TMI2-RA-COR-2023-002 ADMINISTRATIVE CONTROLS 6.7 PRDCEDURES AND PROGRAMS(con't)

1. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM.
2. A. Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of the census, and
3. Participation in an Interlaboratory Comparison Program to ensure that the independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.
6. 8 REPORT[l'iG REOUIREl\ffiNTS RDUTIN"E REPORTS 6.8.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be in accordance .vith 10 CFR 50.4 unless othenvise 1

noted. Some of the reporting requirements of Title 10, Code of Federal Regulations are repeated belov,r. A1'JNUAL RADIOLOGICAL filPlIRONl\ffiNTAL OPERA.TING REPOKT 6.8.1.1 The Annual Radiological Environmental Operating Report covering the operation ofthe unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C ofl.rppendix I to 10 CFR Part 50. ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6.8.1.2 The Annual Radiological Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and (2) in conformance with 10 CFR 50.36a and Section IV.B.l of Appendix I to 10 CFR Part 50. Three Mile Island - Unit 2 6-10 Amendment 48,- 67

Attachment 6 TMI2-RA-COR-2023-002 ADMINISTRATIVE CONTROLS 6.8 REPORTING REOUIREl\4.ENTS (cont'd) BIENJ.'JIAL REPORTS 6.8.1.4 Reports required on a biennial basis shall be submitted on a frequency not to exceed once every hvo years (24 months). The reports shall cover the activities of the unit as described below up to a minimum of6 months prior to the date ofthe filing. Reports required on a biennial basis shall include:

a. All changes made to the PDM:S SAR during the previous update.
b. All changes, tests, or experiments meeting the requirements of 10 CFR 50.59.

SPECIAL REPORTS 6.8.2 Special reports shall be submitted in accordance with 10 CFR 50.4 within the time period specified for each report. NONROUTINE REPORTS 6.8.3 A report shall be submitted in the event that an Exceptional Occurrence as specified in Section 6.13 occurs. The report shall be submitted under one of the report schedules described belov,r. Three Mile Island- Unit 2 6-11 Amendment 4 8,55, 57, 62, 67

Attachment 6 TMI2-RA-COR-2023-002 ADMINISTRATIVE CONTRDLS 6.8 REPORTING REQUIREMENTS (con't) PRDMPT REPORTS 6.8.3.l Those events specified as prompt report occmTences shall be rep01ied within 24 hours by telephone, telegraph, or facsimile transmission to the NRt: followed by a written report to the NRt: within 3 0 days. THIRTY DAY EVEl'H REPORTS 6.8.3.2 Nonroutine events not requiring a prompt report as described in Subsection 6.8.3.1, shall be reported to tho NRG either within 30 days of their occU1Tonco or .vithin tho time limit specified 1 by the reporting requirement of the corresponding certification or permit issued pursuant to Sections 401 or 402 of PL 92 500, the Federal 1,Vater Pollution Control Act (F\VPCA.) i\.mendment of 1972, v1hichever time duration following tho nonroutine e11ent shall result in the earlier submittal. CONTENT OF NillJRDUTINE REPORTS 6.8.3.3 'Nritten 30 day reports and, to the extent possible, the preliminary telephone, telegraph, or facsimile reports shall (a) describe, analyze, and evaluate the occurrence, including eJctent and magnitude of the impact, (b) describe the cause of the occurrence, and (c) indicate the corrective action (including any significant changes made in procedures) taken to preclude repetition of the occurrence and to prevent similar occurrenees imrolving similar eomponents or systems. 6.9 RECORD RETEl'HION 6.9.1 The follov,ring records shall be retained for at least five years:

a. Records of sealed souree and fission detection leak tests and results.
b. Records of annual physical inventory of all sealed source material of record.

6.9.2 The follmving records shall be retained as long as the Licensee has an NRG license to operate or possess the Three J\file Island facility.

a. ALL REPORTABLE EVENTS submitted to the Commission.
b. Records of surveillanee aetivities, inspections and ealibrations required by these Technical Specifications.

Three Mile Island - Unit 2 6-12 Amendment 52, 54, 59, 63, 66, 67

Attachment 6 TMI2-RA-COR-2023-002 ADMINISTRATP/~ CONTRDLS 6.9 RECORDS RETENTION (Con't) C. Records of changes made to tho procedures required by Roeovery Technical Specification 6. 8 .1 and PDMS Technieal Specification 6. 7 .1.

d. Radiation Safety Program Reports and Quarterly Recovery Progress Reports on the March 28, 1979 incident.
o. Records of radioactive shipments.
f. Records and logs of radioactive waste systems operations.
g. Records and drmving changes reflecting facility design modifications made to systems and equipment described in tho Safety 1\nalysis Report, TER, SD, or Safety Evaluation previously submitted to NR£:.
h. Records of new and irradiated fuel inventory, fuel transfers and assembly bumup histories.
l. Records of tra4rnng and qualification for current members of the unit staff.

J- Records of Quality Assurance activities required by the Operating, Recov=ery, or PDMS Quality Assurance Plans.

k. Records of re11iews performed for changes made to procedures or equipment or r011iows of tests and experiments pursuant to 10 CFR 50.59.
1. Records ofmeotings of the PORC and tho GRC, and reports of evaluations prepared by the IOSRD, if applicable to TMI 2.
m. Records of the incident *,vhich occurred on March 28, 1979.
n. Records of unit radiation and contamination surveys.
o. Records of radiation exposure received by all individuals for *.vhom monitoring
             ,vas required.

Three Mile Island - Unit 2 6-13 Amendment48,66,67

Attachment 6 TMI2-RA-COR-2023-002 ADMINISTRi\TIVE CONTROLS 6.9 RECORD RETENTION (Cont'd)

p. Records of gaseous and liquid radioactive material released to the environs.
q. Records of reviews performed for changes made to the OFF SITE DOSE CALCULATION MJ\.NUAL 6.10 RADY-"TION PRDTECTI0NPRDGRAJ\4 Procedures for personnel radiation protection shall be preparnd consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.

&.--1 HIGH RADIATION AREA 6.11.1 In lieu of the "control device" or "alarm. signal" required by paragraph 20.1601 of 10 CFR 20:

a. Each High Radiation Area in which the intensity ofradiation at 30 cm (11 _g in.) is greater than 100 mrem/hr deep dose but less than 1000 mrem/hr shall be barricaded and conspicuously posted as a High Radiation Area, and personnel desiring entrance shall obtain a Radiation Work Permit (RJNP). Any individual or group of individuals entering a High Radiation Area shall (a) use a continuously indicating dose rate monitoring device or (b) use a radiation dose rate integrating device which alarms at a pre set dose le1,i:el (entry into such areas 1.vith this monitoring de1rice may be made after the dose rate le1,i:el in the area has been established and personnel have been made lmov.rledgeable of them),

or (c) assure that a rad4ological control technician provides positive control over activities within the area and periodic radiation surveillance 1.vith a dose rate monitoring instrument.

b. In addition to the requirements of specification 6.11. l .a:
1. Any area accessible to personnel 1.vhere an individual couldreceive in any one hour a defli) dose in e~rness of 1000 mrem at 30 cm (11.g in.) but less than 500 rads at one meter (3.2g ft), from sources of radioactivity shall be locked or guarded to prevent unauthoriwd entry. The keys to these locked barricades shall be maintained under the administrative control of the respective Radiological Controls Supervisor.
2. For individual high radiation areas v,rhere an individual could receive in any one hour a defli) dose in m,cess of 1000 mrem at 20 cm (11.g in.) but less than 500 rads at one meter (3.2g ft.), that are located v,i:ithin large areas such as reactor containment, where no enclosure e~dsts for purposes of locking, and i,vhere no enclosure can be reasonably constructed around the individual area, that indi1ridual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.

The Radiation \Vork Permit is not required by Radiologioal Controls personnel during the performanoe of their assigned radiation protection duties provided they are following radiologioal oontrol prooedures for entry into High Radiation Areas. Three Mile Island - Unit 2 6-14 Amendment48,54,58,66,67

Attachment 6 TMI2-RA-COR-2023-002 ADMINISTRPJIVE CONTROLS Pursuant to 10 CFR Part 20, paragraph 20.160l(c), in lieu of the requirements of paragraph 20.1601(a) and 20.1601(b) of 10 CFR Part 20: 6.11.1 Access to each high radiation area, as defined in 10 CFR 20, in which an individual could receive a deep dose equivalent> 0.1 rem in one hour (at 30 centimeters from the radiation source or from any surface penetrated by the radiation) shall be controlled as described below to prevent unauthorized entry.

a. Each area shall be barricaded and conspicuously posted as a high radiation area.

Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Entrance shall be controlled by requiring issuance of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rate in the immediate work area(s) and other appropriate radiation protection equipment and measures.
c. Individuals qualified in radiation protection procedures or personnel continuously escorted by such individuals may, for the performance of their assigned duties in high radiation areas, be exempt from the preceding requirements for issuance of an R WP or equivalent provided they are otherwise following plant radiation protection procedures for entry into, exit from, and work in such high radiation areas.
d. Each individual or group of individuals permitted to enter such areas shall possess, or be accompanied by, one or more of the following:
1. A radiation monitoring device that continuously indicates the radiation dose rate in the area.
2. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset setpoint is reached. Entry into high radiation areas with this monitoring device may be made after the dose rate in the area has been determined and personnel have been made knowledgeable of it.
3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area.

Three Mile Island - Unit 2 6-14a Amendment48-,g,67

Attachment 6 TMI2-RA-COR-2023-002

4. An individual qualified in radiation protection procedures equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by radiation protection supervision.

6.11.2 In addition to the requirements of Specification 6.11.1, high radiation areas in which an individual could receive a deep dose equivalent> 1.0 rem in one hour (at 30 centimeters from the radiation source or from any surface penetrated by the radiation), but less than 500 rads/hour (at 1 meter from the radiation source or from any surface penetrated by the radiation) shall be provided with a locked or continuously guarded door, or gate, or equivalent to prevent unauthorized entry.

a. The keys to such locked doors or gates, or equivalent, shall be administratively controlled in accordance with a program approved by the radiation protection manager.
b. Doors and gates, or equivalent, shall remain locked except during periods of access by personnel under an approved RWP, or equivalent, to ensure individuals are informed of the dose rate in the immediate work areas prior to entry.
c. Individual high radiation areas in which an individual could receive a deep dose equivalent> 1.0 rem in one hour (at 30 centimeters from the radiation source or from any surface penetrated by the radiation), accessible to personnel, that are located within larger areas where no enclosure exists to enable locking, or that are not continuously guarded, and where no lockable enclosure can be reasonably constructed around the individual area require both of the following access controls:
1. Each area shall be barricaded and conspicuously posted.
2. A flashing light shall be activated as a warning device 6.12 OFFSITE DOSE CALCULATION MANUAL (ODCM)

SUBSTANTIVE CHANGES to the ODCM:

a. Shall be documented and rec0rds of reviews performed shall be reali,a;ed as required by Specification 6.9.2v. This documentation shall contain:
1. Sufficient information to support the change together \Vith the appropriate analyses or evaluations justifying the change(s); and
2. A determination that the change v,rill maintain the level of radioactive effluent control required by 10 CFR 20.1301, 40 CFR Part 190, 10 CFR 50.36a and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
b. Shall become effective after review and acceptance by TMI 2 Solutions, LLC Project Director.
c. Shall be submitted to the Commission in the form ofa complete, legible copy of the entire ODCM as part of or concurrent \Vith the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area ofthe page that was changed, and shall indicate the date (e.g., month/year) the change was implemented. Three Mile Island- Unit 2 6-15 Amendment 48, 54, 55, 58, 67

Attachment 6 TMI2-RA-COR-2023-002 6.13 EXCEPTIO1>JAL OCCURRENCES UNUSUAL OR IIM:PORTANT ENVIRDN1'1ENTAL EVENTS 6.13 .1 Any occurrence of an unusual Of impoftant event that causes Of could potentially cause significant envirnnmental impact causally rnlated 1.vith station opefation shall be FOCOfded and FOpofted to the NRG pef Subsection 6.8.3.1. The following UFO examples of such events: excessive bifd impaction events on cooling towef structuFCs Of meteornlogical tov,zern (i.e., moFC than 100 in any one day); onsite plant of animal disease outbfeaks; unusual mOFtality of any species prntected by the Endangefed Species l'...ct of 1973; fish kills neaF Of downstream of the site. EXCEEDING LIMITS OF RREVANT PERM:ITS 6.13 .2 Any occurrence of exceeding the limits specified in felevant permits and certificates issued by othef Fedefal and State agencies v.11ich arn rnportable to the agency 1.vhich issued the permit shall be repofted to the =NRG in accofdance 1.vith the provisions of Subsection 6.8.3.2. This requirement shall apply only to topics of National Envirnnmental Policy Act concern within the fequifements of the Station NPDES permit as rnlated to TMI 2 dischafges. 6.14 DELETED Three Mile Island - Unit 2 6-15 Amendment 48, 54, 55, 58, 67

Attachment 6 TMI2-RA-COR-2023-002 fJlMINISTRt...TIVE CONTROLS 6.14 (cont'd) DELETED Three Mile Island- Unit 2 6-16 Amendment48,55,67

Attachment 7 TMl2-RA-COR-2023-0002 ATTACHMENT 7 POL and TS Clean Pages contains the clean POL and TS pages. This Attachment supersedes the clean POL and TS pages provided in Reference 1. The current NRC approved TMl-2 Solutions POL and TS are retyped to reflect the addition or deletion of text as identified in the evaluation of proposed changes presented in Attachment 5 and markups presented in Attachment 6. The proposed changes to the TS are considered a major rewrite. Revised formatting (margins, font, tabs, etc.) of content is used to create a continuous electronic file. Revised numbering of pages and sections and the deletion of unused placeholders, where appropriate, is used to condense and reduce the number of pages in the TS without affecting the technical content. Since the changes to the TS are considered a major rewrite, revision bars are not used. 10 Pages Follow

Attachment 7 TMI2-RA-COR-2023-002 TMI-2 SOLUTIONS, LLC DOCKET NO. 50-320 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 2 POSSESSION ONLY LICENSE Amendment No. 67 License No. DPR-73

1. The U.S. Nuclear Regulatory Commission (the NRC or the Commission) has found that:

A. The application for the transfer of the possession only license from Metropolitan Edison Company, Jersey Central Power and Light Company, Pennsylvania Electric Company, and GPU Nuclear, Inc. to TMI-2 Solutions, LLC (the Licensee) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code ofFederal Regulations (10 CFR) Chapter I, and all required notifications to other agencies or bodies have been duly made; B. The facility will be maintained in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission except for those exemptions from specific portions of the regulations, previously granted by the Commission, and still applicable; C. There is reasonable assurance: (i) that the activities authorized by this possession only license can be conducted without endangering the health and safety of the public; and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; D. The licensee is technically qualified to engage in the activities authorized by this possession only license in accordance with the rules and regulations of the Commission; E. The licensee is financially qualified to engage in the activities authorized by this possession only license in accordance with the rules and regulations of the Commission; F. The licensee has satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," of the Commission's regulations; G. The issuance of this possession only license will not be inimical to the common defense and security or to the health and safety of the public; H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental, and other costs and considering available alternatives, the issuance of Possession Only License No. DPR-73 subject to the conditions for protection of the environment set forth herein is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and Possession Only License No. DPR-73 AmendmentNo.67

Attachment 7 TMI2-RA-COR-2023-002 I. The possession of byproduct and special nuclear material and receipt, possession, and use of source material as authorized by the license will be in accordance with the Commission regulations in 10 CFR Parts 30, 40, and 70, including 10 CFR Sections 30.33, 40.32, 70.23, and 70.31.

2. Possession Only License No. DPR-73 is hereby issued to TMI-2 Solutions, LLC to read as follows:

A. This license applies to the Three Mile Island Nuclear Station, Unit 2, (the facility) owned by TMI-2 Solutions, LLC. The facility is located on Three Mile Island in the Susquehanna River in Londonderry Township, Dauphin County, Pennsylvania, about 10 miles southeast of Harrisburg. Prior to entry into Post-Defueling Monitored Storage (PDMS), the facility is described in the Final Safety Analysis Report as supplemented and amended, the various Recovery System Descriptions and Technical Evaluation Reports and the Environmental Report as supplemented and amended. Upon entry into PDMS, the facility is described in the PDMS Safety Analysis Report as supplemented and amended and the Environmental Report as supplemented and amended. B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses: (l)TMI-2 Solutions, LLC, pursuant to Section 103 of the Atomic Energy Act ("Act") and 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," to possess but not operate the facility; (2) TMI-2 Solutions, LLC to possess the facility at the designated location in Dauphin County, Pennsylvania, in accordance with the procedures and limitations set forth in this license; (3) TMI-2 Solutions, LLC, pursuant to the Act and 10 CFRParts 30, 40, and 70, to receive, possess, and use at any time any sealed sources for radiation monitoring equipment calibration; (4) TMI-2 Solutions, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) TMI-2 Solutions, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials which remain at the facility subsequent to the cleanup following the March 28, 1979, accident. The storage ofradioactive materials or radwaste generated at TMI Unit 1 and stored at TMI Unit 2 in accordance with the license for TMI Unit 1 shall not result in a source term that, if released, would exceed that previously analyzed in the PDMS Safety Analysis Report in terms of off-site dose consequences. Possession Only License No. DPR-73 Amendment No. 67

Attachment 7 TMI2-RA-COR-2023-002 C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I, and is subject to all applicable provisions of the Act and to the Commission's rules and regulations, except for those exemptions from specific portions of the regulations granted by the Commission and still applicable, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Technical Specifications The Technical Specifications, as revised through Amendment No. 67 are hereby incorporated into this license. The licensee shall maintain the facility in accordance with the Technical Specifications and all Commission Orders issued subsequent to the date of the possession only license. (2) Physical Protection TMI-2 utilizes a security plan (TMI-2 Materials Security Plan) that is compliant with 10 CFR Part 3 7, "Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material." The plan will implement 10 CFR Part 37 security requirements and also implement specific 10 CFR 73.67 security requirements referenced in the security plan that are not addressed by Part 37. (3) Upon the date of closing, and proceeding until determination of completion of Phase 2 of facility decommissioning, TMI-2 Solutions will maintain a Financial Support Agreement in the amount of$100M, less the value of any cash-funded Provisional Trust Account, Disposal Capacity Easement, and Letter of Credit procured by TMI-2 Solutions for the benefit of the Back-Up Trust Account under the Back-Up & Provisional Trust Agreement. (4) At time of closing, EnergySolutions, Inc. will provide a Parent Guarantee in favor of the FirstEnergy Companies to guarantee the payment and performance of the obligations of TMI-2 Solutions as to the TMI-2 decommissioning. This guarantee makes the resources of EnergySolutions available to help ensure the successful decommissioning of TMI-2, assuring the ability ofTMI-2 Solutions to (i) pay the costs of decommissioning the TMI-2 facility; (ii) protect the public health and safety; and (iii) meet NRC requirements. (5) These financial support conditions (2.C.(3) and 2.C.(4)) may not be voided, canceled, or modified without the prior written consent of the NRC. These financial support conditions are in place and will be maintained as described in the application. The Director of the Office of Nuclear Material Safety and Safeguards shall be informed, in writing, no later than 10 working days after any funds are provided under the terms of the conditions listed above. (6) TMI-2 Solutions, the licensee for TMI-2, will not perform major decommissioning activities which would diminish the historic integrity of the TMI-2 owned and controlled buildings until the applicable historic and cultural Possession Only License No. DPR-73 AmendmentNo.67

Attachment 7 TMI2-RA-COR-2023-002 reviews are completed by the Nuclear Regulatory Commission. D. DELETED E. DELETED F. DELETED G. This license is effective as of the date of issuance and until the Commission notifies the licensee in writing that the license is terminated. FOR THE NUCLEAR REGULATORY COMMISSION (Original signed by Alfred E. Chaffee acting for) Brian K. Grimes, Director Division of Operating Reactor Support Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical Specifications Date of Issuance: Possession Only License No. DPR-73 AmendmentNo.67

Attachment 7 TMI2-RA-COR-2023-002 SECTION 1.0 DEFINITIONS

Attachment 7 TMI2-RA-COR-2023-002 1.0 DEFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications. OFF-SITE DOSE CALCULATION MANUAL 1.12 OFF-SITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of off-site doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the programs required by the Decommissioning Quality Assurance Plan and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by the Decommissioning Quality Assurance Plan and Specifications 6.8.1.2, respectively. Three Mile Island - Unit 2 1-1 Amendment 67

Attachment 7 TMI2-RA-COR-2023-002 Section 2-Not Used Section 3-Not Used Section 4-Not Used Section 5-Not Used

Attachment 7 TMI2-RA-COR-2023-002 SECTION6.0 ADMINISTRATIVE CONTROLS

Attachment 7 TMI2-RA-COR-2023-002 6.0 ADMINISTRATIVE CONTROLS ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6.8.1.2 The Annual Radiological Effluent Release Rep01i covering the operation of the unit during the previous calendar year shall be submitted before May 1 each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and (2) in conformance with 10 CFR 50.36a and Section IV.B. l of Appendix I to 10 CFR Part 50. 6.12 HIGH RADIATION AREA Pursuant to 10 CFR Part 20, paragraph 20.1601(c), in lieu of the requirements of paragraph 20.1601(a) and 20.1601(b) of 10 CFR Part 20: 6.12.1 Access to each high radiation area, as defined in 10 CFR 20, in which an individual could receive a deep dose equivalent> 0.1 rem in one hour (at 30 centimeters from the radiation source or from any surface penetrated by the radiation) shall be controlled as described below to prevent unauthorized entry.

a. Each area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.
b. Entrance shall be controlled by requiring issuance of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rate in the immediate work area(s) and other appropriate radiation protection equipment and measures.
c. Individuals qualified in radiation protection procedures or personnel continuously escorted by such individuals may, for the performance of their assigned duties in high radiation areas, be exempt from the preceding requirements for issuance of an R WP or equivalent provided they are otherwise following plant radiation protection procedures for entry into, exit from, and work in such high radiation areas.
d. Each individual or group of individuals permitted to enter such areas shall possess, or be accompanied by, one or more of the following:
1. A radiation monitoring device that continuously indicates the radiation dose rate in the area.
2. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset setpoint is reached. Entry into high radiation areas with this monitoring device may be made after the dose rate in the area has been determined and personnel have been made knowledgeable ofit.
3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area.

Three Mile Island- Unit 2 6-1 Amendment 67

Attachment 7 TMI2-RA-COR-2023-002

4. An individual qualified in radiation protection procedures equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by radiation protection supervision.

6.12.2 In addition to the requirements of Specification 6.11.1, high radiation areas in which an individual could receive a deep dose equivalent > 1.0 rem in one hour (at 30 centimeters from the radiation source or from any surface penetrated by the radiation), but less than 500 rads/hour (at 1 meter from the radiation source or from any surface penetrated by the radiation) shall be provided with a locked or continuously guarded door, or gate, or equivalent to prevent unauthorized entry.

a. The keys to such locked doors or gates, or equivalent, shall be administratively controlled in accordance with a program approved by the radiation protection manager.
b. Doors and gates, or equivalent, shall remain locked except during periods of access by personnel under an approved R WP, or equivalent, to ensure individuals are informed of the dose rate in the immediate work areas prior to entry.
c. Individual high radiation areas in which an individual could receive a deep dose equivalent> 1.0 rem in one hour (at 30 centimeters from the radiation source or from any surface penetrated by the radiation), accessible to personnel, that are located within larger areas where no enclosure exists to enable locking, or that are not continuously guarded, and where no lockable enclosure can be reasonably constructed around the individual area require both of the following access controls:
1. Each area shall be barricaded and conspicuously posted.
2. A flashing light shall be activated as a warning device.

Three Mile Island - Unit 2 6-2 Amendment 67

Attachment 8 TMl2-RA-COR-2023-0002 ATTACHMENT 8 LIST OF REGULATORY COMMITMENTS The table included in this attachment identifies the Regulatory Commitments in this document. The type of Regulatory Commitment and associated schedule for implementation are provided. Any other statements in this submittal represent intended or planned actions. They are provided for information purposes and are not considered to be Regulatory Commitments. Type Scheduled Completion Date Regulatory Commitment One-Time Continuing Action Compliance Procedures prepared in accordance With the Radiation Protection Program will include controls on work being performed to ensure access and openings of the reactor building X (RB) are only open for the period of Prior to LAR lime necessary to allow passage of implementation material or personnel into or out of ~he RB. Procedures will be developed in accordance with the Radiation Protection Program to perform monitoring and to implement the Prior to LAR source term administrative limits to X implementation ensure potential events do not exceed site boundary dose limits. Each procedure will reference this commitment. For those few areas where the Defueling mix is the controlling factor (A Spent Fuel Pool, A Reactor Coolant Bleed Tank Cubicle and the Prior to LAR Reactor Building 34 7' Elevation in the X implementation vicinity of the A Core Flood Tank), hey will be specifically managed to limit potential combustion sources for DAW source terms. Containers holding DAW will be Prior to LAR closed or sealed from potential X implementation ignition sources prior to movement.}}

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