ML23199A021

TMI-2 Amendment 67 Discrepancies and Revision to SER

ML23199A021
Person / Time
Site:	Three Mile Island
Issue date:	07/31/2023
From:	Amy Snyder Reactor Decommissioning Branch
To:	TMI-2 Solutions
Snyder A
Shared Package
ML23200A187	List: ML23199A021 ML23200A188
References
EPID L-2021-LLA-0038
Download: ML23199A021 (1)

Text

SAFETY EVALUATION BY

THE OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS,

OFFICE OF NUCLEAR REACTOR REGULATION,

AND

OFFICE OF NUCLEAR SECURITY AND INCIDENT RESPONSE

RELATED TO AMENDMENT NO. 67

TO POSSESSION-ONLY LICENSE NO. DPR-73

TMI-2 SOLUTIONS, LLC

THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 2

DOCKET NO. 50-320

REVISION

1.0 INTRODUCTION AND BACKGROUND

The Three Mile Island Nuclear Station (TMINS), located in the Londonderry Township of Dauphin County, is approximately 10 miles southeast of Harrisburg, Pennsylvania. Three Mile Island Nuclear Station, Unit No. 1 (TMl-1) and Three Mile Island Nuclear Station, Unit No. 2 (TMl-2) are located on TMINS. TMINS encompasses approximately 440 acres, and includes adjacent islands to the North, a strip of land on the mainland along the eastern shore of the river, and the area on the eastern shore of Shelley Island that is within the exclusion area (a 2,000-foot radius from a point equidistant between the centers of the TMI-1 and TMI-2 reactor buildings).

On March 28, 1979, TMI-2 experienced an accident initiated by interruption of secondary feedwater flow which led to a core heat up that caused some of the fuel to be damaged. Overall reactor vessel integrity was maintained throughout the accident, but as a result of this accident, small quantities of spent nuclear fuel, damaged core material, and high-level waste (hereinafter collectively referred to as debris material or TMI-2 debris) were transported through the reactor coolant system and the reactor building. In addition, a small quantity of debris material was transported to the auxiliary and fuel handling buildings (AFHB). Further spread of the debris material also occurred as part of the postaccident water processing cleanup activities.1

1 More details about the accident can be found at the U.S. Nuclear Regulatory Commission (NRC) Fact Sheet, Three Mile Island Accident (Agencywide Documents Access and Management System (ADAMS) Accession No. ML082560250). For estimated amounts of debris material, in the form of Uranium Oxide (UO2), that were distributed

Enclosure 2 After the March 28, 1979, accident at TMI-2, the U.S. Nuclear Regulatory Commission (NRC, or Commission) issued an order on July 20, 1979 (ML19208A147), that suspended the authority of the licensee at that time, General Public Utilities Nuclear Corporation (GPUN ), to operate the facility and required GPUN maintain the facility in a shutdown condition in accordance with approved operating and contingency procedures. On August 15, 1988, GPUN submitted a request to amend TMI-2 Operating License No. DPR-73 to a Possession-Only License (POL) and to extensively modify the Technical Specifications (TS) consistent with its plans for long-term storage of the facility (ML20207G801). GPUN called such long-term storage post-defueling monitored storage (PDMS)2, a condition similar to SAFSTOR,3 where the facility would remain until decommissioning at some later time.

In response to GPUNs license amendment request, in August 1989, the NRC staff issued Final Supplement 3 to the Programmatic Environmental Impact Statement Dealing with Post-Defueling Monitored Storage and Subsequent Cleanup (ML20247F778). On April 12, 1990, GPUN informed the NRC staff that it had completed defueling efforts at the TMI -2 facility. The removed fuel (debris material that contains spent fuel) is currently in storage at the U.S.

Department of Energys (DOE) Idaho National Engineering and Environmental Laboratory (INEEL), and the DOE has taken title of the fuel.

On April 25, 1991 (56 FR 19128), the NRC staff published a notice of opportunity for a public hearing regarding the license amendment request for a POL and the proposed changes to the TS (56 FR 19128). On February 20, 1992, the NRC staff issued a safety evaluation (ML20090B852) in which it evaluated the LAR.

On September 14, 1993, the NRC issued Amendment No. 45 to Facility Operating License No DPR-73 for TMI-2 (ML20029E532). This amendment modified Facility Operating License No.

DPR-73 to a possession only license (POL). In a letter dated February 13, 2013, (ADAMS

from the accident and the estimated amount that remain after the initial recovery and cleanup of debris material, refer to the GPU Nuclear, TMI 2 Defueling Completion Report (ML111100641), Section 1 and 2 and Table 2-1.

2 Because of the 1979 accident at TMI-2, GPUN did not follow the standard process for cessation of operations provided in Section 50.82 of Title 10 of the Code of Federal Regulations (10 CFR), Termination of license. On August 16, 1988, GPUN proposed placing the facility in a storage mode after the completion of the defueling process to allow decay of the radionuclides remaining in the facility. GPUN proposed the near-term implementation of PDMS without further preparation for decommissioning. By postponing preparation for decommissioning, the final cumulative dose total would be reduced through natural decay of the radioactive material in the plant and the expected advances in cleanup technology. An evaluation included in Supplement No. 3 of the Programmatic Environmental Impact Statement (PEIS) (NUREG-0683, Oct. 1984) estimated the dose savings to range from 3600 to 9100 person-rem.

The PDMS Safety Analysis Report (Aug. 16, 1988) estimated the dose savings to range from 4500 to 9800 person-rem. The large dose savings due to PDMS significantly reduced the overall occupational dose total for TMl-2 (ML15278A310).

3 As explained in the Backgrounder on Decommissioning Nuclear Power (ML040340625), DECON and SAFSTOR are two different decommissioning strategies. Under DECON (immediate dismantling), soon after the nuclear facility closes, equipment, structures, and portions of the facility containing radioactive contaminants are removed or decontaminated to a level that permits release of the property and termination of the NRC license. Under SAFSTOR, often considered deferred dismantling, a nuclear facility is maintained and monitored in a condition that allows the radioactivity to decay, or for other reasons, such as the availability of waste disposal sites. Afterwards, the plant is dismantled, and the property decontaminated. A licensee may also combine the two strategies by dismantling and decontaminating some portions of the facility while leaving other parts in SAFSTOR. PDMS is a form of SAFSTOR.

There is no formal declaration of a strategy: A facility is said to be in DECON when active decommissioning work is underway. Decommissioning must be completed within 60 years of the plant ceasing operations. A time beyond that would be considered only when necessary to protect public health and safety in accordance with NRC regulations.

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Accession No. ML12349A291), the NRC stated that September 14, 1993, is considered the date of TMl-2's cessation of operations (ML20029E532).

Although the POL amendment was issued on September 14, 1993, the TS in amendment No.

45 were not compatible with PDMS. The PDMS TS could not be implemented until the final phases of the current cleanup effort were completed, the NRC staff had verified the implementation of the PDMS requirements and commitments, and GPUN had satisfied the PDMS license conditions. Therefore, the purpose of the licensing action for Amendment No. 48 was to issue the PDMS TS after the PDMS requirements and commitments were satisfied and all the license conditions were met. Documents submitted by GPUN related to Amendment No.

48 were submitted on January 18 (ML20127M275), May 26 (ML20036B962), October 24 (ML20059F040), and November 12, 1993 (ML20059M207).

Amendment No. 48 (ML20059D157 (Pkg)), effective December 28, 1993, replaced the TMI-2 Appendix A and B TS with the PDMS TS to facilitate long-term monitored storage of the facility.

The Commission's related evaluation of the amendment is contained in a safety evaluation dated December 28, 1993 (ML20059D177).

On December 18, 2020, the POL was transferred to TMI -2 Solutions, LLC (TMI-2 Solutions, or the Licensee). Following the license transfer (ML20352A381) and closing (ML20350B569), TMl-2 entered what the Licensee called Phase 1 of its decommissioning plan while the facility remains in a PDMS condition.

All spent fuel, including debris material that contains spent fuel, has been removed, except for some debris material in the reactor coolant system and ex-vessel4 (a total quantity of residual fuel is estimated to be less than 1125 kg or approximately 1 percent of the original core inventory). This debris material is primarily in the form of finely divided, small particle-size sediment material, resolidified material either tightly adherent to components or in areas inaccessible to defueling, and adherent films on surfaces contained within piping, tanks, and other components.

1.1 TMI-2 Solutions License Amendment Request

By letter dated February 21, 2021(ADAMS Package Accession No. ML21057A047, as supplemented on May 5 (ML21133A264); January 7, 2022 (ML22013A177), March 23, (ML22101A079), April 7, (ML22101A077), April 8 (ML22105A092), May 16, (ML22138A285),

September 29 (ML22276A024), October 29, (ML22307A082) and December 30, 2022 (ML22364A194); and January 27 (ML23033A103), February 14 (ML23049A004), and March 30, 2023 (ML23090A216), TMI-2 Solutions submitted a license amendment request (LAR) seeking to revise the POL and the associated TS to support TMI-2s transition from the PDMS condition to that of a facility undergoing radiological decommissioning (DECON) pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.82(a)(7).

Specifically, the LAR seeks to eliminate certain TS, limiting conditions for PDMS, and surveillance requirements that have already been met, or that are no longer applicable based the facilitys current radiological conditions. The LAR also proposes to relocate administrative controls from Section 6, Administrative Controls, to the Decommissioning Quality Assurance

4 Ex-vessel, means the areas outside the reactor vessel (RV) that resulted from fuel material transport within the RV and from the RV to locations outside the RV as described in the GPU Nuclear, TMI 2 Defueling Completion Report (ML111100641).

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Plan (DQAP), subsequently controlling them in accordance with 10 CFR 50.54(a) pursuant to the criteria contained in 10 CFR 50.36 and in accordance with the recommendations, guidance and purpose of NRC Administrative Letter 95-06 (ML20101P963). Finally, the LAR also notifies the NRC of an update to the facilitys Safe Fuel Mass Limits (SFML).

2.0 REGULATORY EVALUATION

This safety evaluation assesses the acceptability of the proposed TMI -2 Solutions LAR, as supplemented, to revise the POL and the associated TSs to support TMI-2s transition from the PDMS condition to that of a facility undergoing DECON. The regulatory requirements and associated guidance on which the NRC bases evaluation of this LAR is detailed below.

2.1 Applicable Regulations and Guidance

10 CFR 50.82, Termination of license.

By letter dated August 14, 2012, (ML12235A227) GPUN informed the NRC of the status of TMl-2 relative to the 1996 Decommissioning Rule changes, specifically related to 10 CFR 50.51, Continuation of license, and 10 CFR 50.82, Termination of license. The letter stated the intent to submit a Post-Shutdown Decommissioning Activities Report (PSDAR) that describes the planned decommissioning activities, schedule, cost estimates, and the environmental impacts of TMl-2 facility specific decommissioning. GPUN submitted the TMI-2 PSDAR, Rev. 0 under 10 CFR 50.82(a)(4) on June 28, 2013 (ML13190A366).

10 CFR 50.36, Technical Specifications.

Under 10 CFR 50.36(e), the provisions of 10 CFR 50.36 apply to each nuclear reactor licensee whose authority to operate the reactor has been removed by license amendment, order, or regulation. 10 CFR 50.36(c)(6) states:

Decommissioning. This paragraph applies only to nuclear power reactor facilities that have submitted the certifications required by § 50.82(a)(1) [5] and to non-power reactor facilities which are not authorized to operate. TS involving safety limits, limiting safety system settings, and limiting control system settings; limiting conditions for operation; surveillance requirements; design features; and administrative controls will be developed on a case -by-case basis.

Under 10 CFR 50.92(a), determinations on whether to grant an applied-for license amendment are to be guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Both the common standards for licenses in 10 CF R 50.40(a)

(regarding, among other things, consideration of the operating procedures, the facility and equipment, the use of the facility, and other TS, or the proposals) and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable

5 Pursuant to 10 CFR 50.82(a)(1)(i), [w]hen a licensee has determined to permanently cease operations the licensee shall, within 30 days, submit a written certification to the NRC. Per 10 CFR 50.82(a)(1)(ii), [o]nce fuel has been permanently removed from the reactor vessel, the licensee shall submit a written certification to the NRC. Per 10 CFR 50.82(a)(2), [u]pon docketing of the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel,..., the 10 CFR part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel.

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assurance that the activities at issue will not endanger the health and safety of the public, and that the applicant will comply with the Commissions regulations.

Therefore, when deciding whether to amend the TS for a permanently shut down and defueled reactor such as TMI-2, the NRC staff considers, on a case-by-case basis, whether the proposed amended TS, along with the operating procedures, the facility and equipment, and the use of the facility collectively provide reasonable assurance that the applicant will comply with the Commission's regulations, and that the health and safety of the public will be protected.

10 CFR 50.48, Fire Protection.

NRC regulations in 10 CFR 50.48(f) state, in part, that licensees that have submitted the certifications required under 10 CFR 50.82(a)(1) maintain a fire protection program to address the potential for fires that could cause the release or spread of radioactive materials.

In 10 CFR 50.48(f), the NRC established the requirement for maintaining a fire protection program once a licensee has submitted the certifications required under 10 CFR 50.82(a)(1).

10 CFR 50.51, Continuation of license.

10 CFR 50.51 (b) states:

(b) Each license for a facility that has permanently ceased operations, continues in effect beyond the expiration date to authorize ownership and possession of the production or utilization facility, until the NRC notifies the licensee in writing that the license is terminated. During such period of continued effectiveness, the licensee shall---

(1) Take actions necessary to decommission and decontaminate the facility and continue to maintain the facility, including, where applicable, the storage, control, and maintenance of the spent fuel, in a safe condition, and; (2) Conduct activities in accordance with all other restrictions applicable to the facility in accordance with the NRC regulations and the provisions of the specific 10 CFR Part 50 license for the facility.

10 CFR 50.2, Definitions.

10 CFR 50.2 states that Safety-related structures, systems and components means those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in § 50.34(a)(1) or§ 100.11 of this chapter, as applicable.

Major decommissioning activity means, for a nuclear power reactor facility, any activity that results in permanent removal of major radioactive components, permanently modifies the

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structure of the containment, or results in dismantling components for shipment containing greater than Class C waste in accordance with § 61.55 of this chapter.

10 CFR 20, Standards for Protection Against Radiation

The staff reviewed these regulations when reviewing then proposed change to TS 6.11, High Radiation Areas. Also, the staff used Regulatory Guide 8.38. Rev. 1, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants. Regulatory Guide (RG) 8.38, published in May 2006 describes methods that the NRC considers acceptable for complying with the regulations regarding radiation safety and protection and for specifically complying with implementing the requirements of 10 CFR 20 Sections 1101, 2101, 1601, and 1602.

10 CFR 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions

The NRC staff has evaluated the potential environmental impacts associated with the proposed action to address the requirements of 10 CFR Part 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions."

The NRC staff used guidance found in NUREG-1748, Environmental Review Guidance for Licensing Actions Associated with NMSS Programs (ML032450279).

3.0 TECHNICAL EVALUATION

The NRC staff has reviewed the Licensees regulatory and technical analyses in support of its proposed changes, as described in the LAR dated February 21, 2021, as supplemented. The proposed amendment would revise the POL and the associated TS to support the transition of TMI-2 from a PDMS condition to that of a facility undergoing radiological decommissioning (DECON) with all TMI-2 spent fuel removed from theTMI-2 site with remaining debris material in the TMI-2 reactor pursuant 10 CFR 50.82(a)(7).

3.1 Proposed License Changes

3.1.1 License Condition 2.C.(1)

TMI-2 Solutions proposes to modify License Condition 2.C.(1) to update the amendment number in the condition, as well as to update the language to refer to maintaining the facility and instead of operating the facility.

The Licensee states in its LAR that the revisions will reflect approval of the proposed license change. The Licensee explains that the TMl-2 POL no longer authorizes use of the facility for power operation or emplacement or retention of fuel into the RV as provided in 10 CFR 50.82(a)(2). Therefore, the removal of the reference to an operating plant provides clarity in the 10 CFR Part 50 license description, and the change is consistent with the requirements associated with a decommissioning facility.

The NRC staff reviewed the TS definitions proposed to reflect approval of the amendment and to reflect the authorized use of the facility, and agrees that the change in term from operate to maintain is relevant to TMI-2 in that the Licensee is no longer authorized to produce power or emplace or retain fuel in its RV. Further, updating the POL amendment number would be necessary upon approval of the LAR and would be administrative in nature ; however, the next

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amendment that would be issued would be no. 67, not 65. The NRC staff finds the modifications to the License Condition 2.C.(1) acceptable and administrative in nature.

3.1.2 License Condition 2.D

GPUN was required by the License Condition 2.D, Special Auxiliary and FHB Ventilation Study, to complete a special monitoring program of the auxiliary and fuel handling buildings (AFHB) ventilation systems airborne levels prior to terminating continuous operation of the AFHB. In the LAR, the Licensee proposed to delete License Condition 2.D because it states that the requirements of the license condition have been satisfied.

The NRC staff reviewed the Licensees proposed deletion of License Condition 2.D and noted that the information has already been submitted to the NRC. The NRC staff confirmed that the NRC has evaluated the information submitted (ML20084L576) and found it acceptable in an SE dated December 28, 1993 (ML20062K136). Therefore, the NRC staff finds that because the Licensee has satisfied the requirements of License Condition 2.D, there is no regulatory basis for the Licensee to maintain License Condition 2.D in the POL, so it can be deleted. Deleting this license condition is administrative in nature.

3.1.3 License Condition 2.E

GPUN was required by License Condition 2.E, Unfiltered Leak Rate Test, to develop NRC approved surveillance requirements for the Reactor Building unfiltered leak rate test that will be incorporated as Section 4.1.1.2 of the PDMS TS. In the LAR, TMI-2 Solutions proposed to delete License Condition 2.E because it states that the requirements of the license condition have been satisfied.

The NRC staff reviewed the proposed deletion of License Condition 2.E and found that the information was already submitted for the PDMS surveillance status via Amendment No. 16 to the GPUN PDMS SAR, dated January 18, 1993 ( ML20237F143) and evaluated by the NRC in an SE dated, December 28, 1993 (ML20059D177). Therefore, the NRC staff finds that the requirements of License Condition 2.E have been met and is no longer needed. The action is administrative in nature.

3.1.4 License Condition 2.F

GPUN was required through License Condition 2.F to provide additional submittals prior to PDMS. Specifically, GPUN had to submit and implement a site Flood Protection Plan, a site Radiation Protection Plan, an ODCM, a PDMS Fire Protection Program Evaluation, a PDMS Quality Assurance Plan and a Radiological Environmental Monitoring Plan. Additionally, GPUN had to submit to the NRC the results of the completed plant radiation and contamination surveys prior to entry into PDMS. TMI-2 Solutions proposes to delete License Condition 2.F in its entirety because it states that the requirements of the condition have been satisfied. The Licensee states that the information was submitted and the NRC evaluated the information and found it acceptable as documented in SER "Post-Defueling Monitored Storage" dated December 28, 1993 (ML20059D177).

The NRC staff confirmed that the completed plant radiation and contamination surveys results prior to entry into PDMS were submitted by GPUN by letter dated November 12, 1993 (ML20059M213). The NRC staff confirmed that all the required information was submitted by

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GPUN, and the NRC found it acceptable. Therefore, the requirements of this License Condition have been met and it can now be deleted. The change is administrative in nature.

3.1.5 POL Enclosure Statement

The Licensee proposes that the enclosure to the TS be modified to delete the statement that indicates that there is an Appendix B to the TS. TMI-2 Solutions states that there is no Appendix B associated with the TMl-2 TS, therefore this text is proposed for deletion.

The NRC staff agrees with this proposed change because Appendix B, Environmental Technical Specifications, is no longer applicable. It was removed via license Amendment No.

48 when Appendices A and B where combined into Appendix A (ML20059D157 (Pkg)). When License Amendment No. 65 was issued (ML22189A196), the NRC staff inadvertently issued an amendment stating Appendices A & B.

However, the NRC staff noted the title page to the TS (ML20059D157 (Pkg)) identifies the TS as Technical Specifications for PDMS and not Appendix A. Therefore, the NRC staff finds the enclosure statement modification should be made as follows:

Enclosure:

Technical Specifications

And the title page be modified to: Decommissioning Technical Specifications.

The NRC concludes that the changes are administrative in nature.

3.2 Proposed Technical Specification (TS) Changes

In its application, TMl-2 Solutions committed to continue to conduct activities in accordance with the definitions in 10 CFR 50.2.

Further, in the Licensees application, the Licensee states that the proposed changes are consistent with the intent of the license and accompanying TS issued to the following facilities that have been permanently shutdown and defueled and cites the following amendment applications that were approved:

(1) Vermont Yankee Nuclear Power Station, for which an amendment was issued on October 7, 2015 (ML15117A551); (2) Kewaunee Power Station, for which an amendment was issued on February 13, 2015 (ML14237A045); (3) San Onofre Nuclear Generating Station, Units 2 and 3, for which an amendment was issued on July 17, 2015 (ML15139A390); and (4) Crystal River Nuclear Plant, Unit 3, for which an amendment was issued on September 4, 2015 (ML15224B286).

TMl-2 Solutions states in its application that it has evaluated the LARs identified above and associated NRC issued amendments and safety evaluation reports from the perspective of ensuring that proposed revisions to, and deletions of, TS from the TMl-2 license are similar with those submitted by other licensees and accepted by the NRC.

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3.2.1 TS Section 1.0, Definitions

The Licensee proposed to relocate the following definitions from the TS to the Decommissioning Quality Assurance Program (DQAP):

• OFF-SITE DOSE CALCULATION

• SUBSTANTIVE CHANGES

• SITE BOUNDARY

• NPDES PERMIT

The Licensee proposed to revise the definition for OFF-SITE DOSE CALCULATION MANUAL by revising it to remove references related to programs in Section 6.7.4, the Annual Radiological Environment Operating Report, and TS 6.8.1.3 and relocating the definition to the DQAP.

The Licensee indicated that these definitions are cited in TS that are proposed to be relocated to the DQAP, and therefore the definitions are proposed to be relocated.

The Licensee proposed to delete the following definitions from the TS:

• ACTION

• OPERABLE - OPERABILITY

• CHANNEL CALIBRATION

• CHANNEL CHECK

• CHANNEL FUNCTIONAL TEST

• FREQUENCY NOTATION

• CONTAINMENT ISOLATION

• BATCH RELEASE

• CONTINUOUS RELEASE

• REPORTABLE EVENTS

• STAGGERED TEST BASIS

• MEMBER(S) OF THE PUBLIC

• UNRESTRICTED AREA

• FREQUENCY NOTATION

The Licensee indicated that these definitions are proposed for deletion since the terms are either defined in the ODCM, are not used in any TS, are codified in the applicable regulations (e.g., 10 CFR 50.72 and 10 CFR 50.73 specific to reportable events), or are defined in regulations (e.g., 10 CFR 20.1003).

The NRC staff reviewed the TS definitions proposed for relocation as well as the revision to the definition discussed above and concludes that the Licensees proposed revision accurately reflects the permanent shutdown and defueled condition and that those terms proposed for relocation listed above will continue to apply to the facility and will reside in the DQAP along with the associated TS. The NRC staff reviewed the TS definitions proposed for deletion and concludes that the inclusion of those terms in the TS would no longer have any regulatory basis after PDMS. Therefore, the NRC staff finds the revision, relocation, and deletion of the definitions from the above TS are acceptable and administrative in nature.

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3.2.2 TS Section 2.0, Safety Limits

Pursuant to 10 CFR 50.36(c)(1), safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. In its LAR, TMI-2 Solutions proposes to delete TS Section 2.0 in its entirety because there are no safety limits associated with TMl-2 because it does not have a reactor coolant pressure boundary and the facility is permanently defueled. The NRC staff agrees with the Licensee; accordingly, there is no regulatory basis for TMl-2s POL to contain TS Section 2.0, Safety Limits. The NRC staff concludes that deletion of TS Section 2.0 in its entirety is appropriate, acceptable, and is administrative in nature.

3.2.3 TS Section 3/4.0, Limiting Conditions for PDMS and Surveillance Requirements

TS Section 3/4.0 contains the Limiting Conditions (LCs) for PDMS and associated Surveillance Requirements for TMl-2. TMl-2 Solutions proposes to delete all PDMS-related TS and not establish any new TS applicable for use in Phase 1b and Phase 2 of decommissioning (i.e.,

DECON).

In its LAR, the Licensee explains that the PDMS LCs and TS 3/4.0 LCs for PDMS associated surveillance requirements presented in TS Section and Surveillance Requirements 3/4.0 are applicable in Phase 1a. The Licensee proposes to delete these in their entirety because these TS, LCs, and surveillance requirements only apply to the facility in a PDMS condition. For these same reasons, the Licensee proposes to delete TS Sections 3/4.1 through 3/4.4. The Licensee accordingly concludes that there is no need to define LCs for decommissioning in Phase 1b and Phase 2 (i.e., DECON) or define any associated surveillance requirements.

The NRC staff evaluated the Licensees proposal and agrees that TS Section 3/4.0 should be deleted in its entirety because these requirements apply to PDMS. Also, the NRC staff concludes that no LCs for the facility in DECON or associated surveillance requirements would be needed if TS Sections 3/4.1 through 3/4.4 are deleted. The NRC staffs evaluation below concludes that TS Sections 3/4.1 through 3/4.4 may be deleted. Therefore, the NRC staff finds it acceptable to delete TS Section 3/4.0- LCs for PDMS and Surveillance Requirements and is administrative in nature.

3.2.4 TS Section 3/4.1, Containment Systems

TS Section 3/4.1 contains LCs for PDMS that assures that containment is maintained as a barrier for the residual contamination which remains inside the containment. TS Section 3/4.1 includes TS 3.1.1.1 Containment Isolation, TS 3.1.1.2 Unfiltered Leak Rate Testing, and TS 3.1.1.3 Containment Air Locks. In its February 19, 2021, LAR, TMI-2 Solutions proposed to delete TS Section 3/4.1-Containment Systems because these TS do not satisfy any of the four requirements established in 10 CFR 50.36(c)(2)(ii)(A)-(D). The NRC staffs evaluation of the satisfaction of the four requirements is discussed below in further detail.6

Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

6 Additionally, the Licensee indicated that the control of residual contamination inside the containment will be accomplished by the RPP and through implementation of procedures.

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Licensees Evaluation

The Licensee indicates that TMl-2 does not have a reactor coolant pressure boundary.

Accordingly, the Licensee concluded that Criterion 1 is not applicable to TMl-2.

Staff Evaluation The NRC staff evaluated the LAR and finds that since TMI-2 no longer has a reactor coolant pressure boundary, TS Section 3/4.1 no longer meets Criterion 1.

Criterion 2: 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.

Licensees Evaluation

The Licensee indicated that TMI-2 has removed 99 percent of its fuel and that currently there are no DBAs or transients associated with TMl-2. The Licensee concluded that Criterion 2 is not applicable to TMl-2.

Staff Evaluation

The NRC staff evaluated the information provided in the Licensees submittal and finds that TMI-2 is no longer licensed for power operations, 99 percent of its fuel has been removed from the facility and is permanently defueled, and that there are no DBAs or transients associated with the facility. The NRC staff noted that TMI-2 Solutions performed a containment fire analysis, and it does not credit containment closure or isolation in its mitigation. Therefore, the NRC staff finds that TS Section 3/4.1 no longer meets Criterion 2.

Criterion 3: A structure, system, or component 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.

Licensees Evaluation

The Licensee indicated that there are no DBAs or transients associated with TMl-2. The Licensee concluded that Criterion 3 is not applicable to TMI-2.

NRC Staff evaluation

The NRC staff evaluated the information provided in the LAR and finds that TMI-2 is has a POL and is permanently defueled and that there are no DBAs or transients associated with the facility. The staff noted that TMI-2 Solutions performed a containment fire analysis, and it does not credit containment closure or isolation in its mitigation.

Therefore, the NRC staff finds that TS Section 3/4.1 no longer meet s Criterion 3.

Criterion 4: A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

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Licensees evaluation of Criterion 4

TMI-2 Solutions indicated that TMI-2 has removed 99 percent of its fuel and that currently there are no DBAs or transients associated with TMl-2. The Licensee concluded that Criterion 4 is not applicable to TMl-2.

NRC Staff evaluation

The NRC staff evaluated the information provided in the LAR and finds that TMI-2 has a POL, is permanently defueled and that there are no DBAs or transients associated with the site. This has reduced the risk of an accident capable of significantly impacting the public health and safety. The staff also notes that the applicant performed a containment fire analysis (which is not considered a DBA), and the mitigation of this event does not credit containment closure or isolation. The consequences of the containment fire are bounded by the results of the update to the containment fire scenario fire analysis (ML23033A103), and remain below regulatory limits. The NRC staff has not identified any operational experience that would require crediting TMI-2 containment systems for the protection of public health and safety. The NRC staff finds that the containment systems are not significant for the protection of public health and safety, therefore, TS Section 3/4.1 no longer meets Criterion 4.

The NRC staff evaluated the criterion identified in 10 CFR 50.36(c)(2)(ii)(A)-(D) and determined that TS 3.1.1.1, TS 3.1.1.2, and TS 3.1.1.3 no longer meet the criterion for inclusion in the TS and therefore the NRC staff finds it acceptable to delete these TS from the TMI-2 POL. The NRC staff concludes that these changes are administrative in nature.

3.2.5 TS Section 3/4.2, Reactor Vessel Fuel

TS Section 3/4.2 contains LCs for PDMS to assure that no more than 42 kg of fuel (i.e., UO2) may be removed from the 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 (ML111100641) and the criticality safety analyses. TMI-2 proposed to delete this LC in its entirety because, according to the licensee, a criticality event is not credible.

As explained by TMI-2 Solutions, the current SFML in the RV was determined to be 93 kg of core debris material. Based on past industry practice, a limit of approximately 45 percent of the SFML was placed on the amount of debris material that may be removed from the RV or rearranged in the RV, thus resulting in the current LC that no more than 42 kg of fuel is allowed to be removed or rearranged without NRC approval. Calculating the SFML in this manner is intended to ensure subcriticality even in the case of dual errors. The Licensees LAR, specifically, Calculation TMl2-EN-RPT-0001 "Determination of the Safe Fuel Mass Limit for Decommissioning TMl-2," Attachment 5, as updated through Attachment 1 to TMI-2 Solutions submittal dated April 7, 2022 (ML22101A077), provides a proposed basis to increase the TMI-2 SFML from 42 kg to 1200 kg. According to TMI-2 Solutions, this calculation demonstrates that the remaining core debris material does not present a credible criticality concern. Thus, TMI-2 Solutions proposes that TS Section 3/4.2 is not necessary after PDMS and can be deleted in its entirety. The Licensee states that the TS does not apply to TMl-2 while in DECON.

An estimated 1097 kilograms (kg) of residual UO 2, ~1.2 percent of original TMI-2 inventory, is still present within the RV and in various locations outside of the RV. To support the defueling of

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the remaining fissionable material, TMI-2 Solutions established an updated SFML based on more realistic conditions and sampling data. The new SFML intends to bound all remaining activities including removal of debris material from the RV, movement to the reactor cavity or other area intended for segmentation, and movement to loading of the transportable storage container (TSC).

The remaining debris material is in the form of finely divided, small particle-size sediment material; resolidified material either tightly adherent to components or in areas inaccessible to defueling at the time of the initial clean up after the 1979 accident up until the completion of the defueling7; and adherent films on surfaces contained within piping, tanks, and other components. The debris material is present both inside and outside of the RV, with most of the mass residing in the lower head of the RV. Licensee evaluations of the residual fuel suggests that no discrete (i.e., neutronically de-coupled) location has in excess of 127 kg UO2, and the total estimate of mass present outside of the RV is 170 kg UO2.

The NRC staff determined that the Licensees associated activities that are to be conducted via the 10 CFR 50.59 process involving the debris material do not present any credible criticality hazards. Therefore, the NRC staff determined that the Licensee has described a program that will provide reasonable assurance of adequate protection for the health and safety of workers and the public against credible criticality hazards related to the Licensees proposed activities, and that the associated activities that are to be conducted via the 10 CFR 50.59 process involving the debris material will be conducted such that subcriticality is assured under normal and all credible abnormal conditions. Therefore, the NRC staff finds it acceptable to delete TS 3/4.2 from the POL in its entirety.

3.2.5.1 Material Control and Accounting

Because of the proposed deletion of the SFML TS establishing a LC for PDMS, the NRC staff also evaluated the Licensees material control and accounting (MC&A) practices for the debris material in order to ensure that the proposed changes and associated activities that are to be conducted via the 10 CFR 50.59 process involving the remaining debris material are compliant with MC&A requirements, as modified by the 1985 MC&A exemption (ML20138D392).

The MC&A requirements applicable to the Licensee are contained in 10 CFR Part 74, Subpart B, General Reporting and Recordkeeping Requirements. Licensees who possess, transfer, or receive SNM in a quantity of one gram or more of contained uranium-235, uranium-233, or plutonium are subject to the general reporting and recordkeeping requirements of 10 CFR 74.11, 74.13, 74.15, and 74.19.

By letter dated July 29, 2022 (ADAMS package No. ML22210A080), the NRC sent the Licensee a RAI which included 7 items related to MC&A. Responses to the RAIs from TMI-2 Solutions were received by the NRC by letter on September 29, 2022 (ML22276A024). A clarification call between the NRC and the Licensee concerning two minor RAI MC&A issues was conducted on November 16, 2022. An email summarizing the call was sent to the Licensee on November 16, 2022 (ML22321A007). By letter dated January 27, 2023 (ML23033A103), TMI-2 Solutions submitted to the NRC a revision of the MC&A RAI responses addressing the clarification issues.

Additionally, the licensee voluntarily agreed to the addition of a TS (TS 6.16) requiring the licensee to conduct MC&A activities to address applicable regulatory requirements, including, 10 CFR 74.11, 74.13, 74.15, 74.19(a)(1)-(4), 74.19(b), and 74.19(c); however the scope of

7 See TMI Nuclear Station Unit 2 Defueling Completion Report (Ref.3)

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74.19 (a)(1) and 74.19(c) are as described in Licensee response to RAI 14 and RAI 16 (ML22276A024).

Reports of loss or theft or attempted theft (10 CFR 74.11)

The requirements at 10 CFR 74.11(a) states that each licensee who possesses 1 gram or more of contained uranium-235, uranium-233, or plutonium is to notify the NRC Operations Center within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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.

In the response to RAI 11 (ML22276A024), TMI-2 Solutions describes its plan for reporting within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> any loss, theft, unlawful diversion, or attempted theft or unlawful diversion, of SNM. The Licensee states that personnel access is controlled, random searches are conducted, and security personnel are notified upon detection of loss, theft, or unlawful diversion, or attempted theft or unlawful diversion, of SNM. The Licensee also states that the details for meeting the one-hour reporting requirement of 10 CFR 74.11(a) are included in procedure TMI2-RA-PR-0005, Reporting of Events and Conditions. The Licensee affirms that activities are performed to ensure any indicator of loss, theft or diversion, or attempted theft or diversion, of SNM is responded to the NRC in a timely manner in accordance with 10 CFR 74.11(a).

The NRC staff reviewed the Licensees description for notifying the NRC of loss, theft or unlawful diversion, or an attempt at theft or unlawful diversion, of SNM. Based on the review, the NRC staff determined that the Licensees MC&A measures include adequate procedures to ensure that the NRC is notified in a timely manner in the event of a loss, theft or unlawful diversion, or attempted theft or unlawful diversion, of SNM. As agreed to by the Licensee (ML23090A214), an additional TS under TS 6.0, Administrative Controls, has been added to capture the licensees commitments in its licensing basis. Therefore, the NRC staff finds that the Licensee meets the requirement of 10 CFR 74.11.

TS 6.16.1

Establish written procedures to address applicable MC&A activities in 10 CFR 74.11, 74.13, 74.15, 74.19(a)(1) through (4), 74.19(b),and 74.19(c); however, for 74.19(a)(1) and 74.19(c), the scope is as described in Licensee response to RAI 14 and RAI 16 (ML22276A024).

Material Status Reports (10 CFR 74.13)

The requirements at 10 CFR 74.13(a) states that each licensee possessing SNM in a quantity totaling one gram or more of contained uranium -235, uranium-233, or plutonium must 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.

In the response to RAI 12 (ML22276A024), TMI-2 Solutions refers to an exemption granted by the NRC in 1985 for submitting material balance reports and physical inventory listings, which expired upon the completion of defueling and offsite shipment of packaged fuel debris (ML20138D392). At that time, material balance reports and physical inventory listings were required by 10 CFR 70.53. In a 2002 NRC rulemaking, 10 CFR 70.53 requirements were moved into Part 74 as 10 CFR 74.13 and 10 CFR 74.17. Thus, the 1985 exemption from 10 CFR 70.53 is equivalent to an exemption today from 10 CFR 74.13 and 10 CFR 74.17. Since 10 CFR 74.17

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does not apply to TMI-2, as it is not subject to Subparts C, D, or E of 10 CFR Part 74; the exemption only applied to 10 CFR 74.13. The Licensee states that the final SNM accounting will reconcile the total SNM shipped and the SNM loaded into dry storage canisters to the final status survey, and that the final material balance reports will be generated at that time by written procedure.

The NRC reviewed the licensees description of preparing and submitting material status reports. The NRC determined that the 1985 exemption from 10 CFR 70.53 provides the equivalent exemption from 10 CFR 74.13. Based on the review, the NRC staff determined that the Licensees MC&A measures include adequate accounting practices so that the final SNM accounting will be able to adequately reconciles the total SNM shipped and the SNM loaded into dry storage canisters to the final status survey. As agreed to by the Licensee (ML23094A060),

a TS under TS 6.0, Administrative Controls, was added to capture this procedure as part of the licensing basis (see TS 6.16.1 below). Therefore, the NRC staff finds that the Licensee appropriately addresses the requirement of 10 CFR 74.13(a).

TS 6.16.1

Establish written procedures to address applicable MC&A activities in 10 CFR 74.11, 74.13, 74.15, 74.19(a)(1) through (4), 74.19(b),and 74.19(c); however, for 74.19(a)(1) and 74.19(c), the scope is as described in Licensee response to RAI 14 and 16 (ML22276A024).

Nuclear Material Transaction Reports (10 CFR 74.15)

The requirements at 10 CFR 74.15 states that each licensee who transfers or receives SNM in a quantity of 1 gram or more of contained uranium-235, uranium-233, or plutonium is 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.

In the response to RAI 10 (ML22276A024 and ML23033A103), the Licensee states that material transaction reports will be created in accordance with written procedures for all packages containing 1 gram or more of SNM shipped from TMI-2. In the response to RAI 13, the Licensee states that characterization prior to shipment of low-level radioactive waste packages will meet 10 CFR 74.15(a) requirements. The Licensee states that for SNM in dry cask storage, characterization will meet 10 CFR 74.15(a) requirements after completion of the final status survey per written procedure. The Licensee affirms that material transaction reports will be submitted at the required frequencies.

The NRC staff reviewed the Licensees description of completing and submitting nuclear material transaction reports. Based on the review, the NRC staff determined that the Licensees MC&A measures include adequate procedures to ensure that transfers and receipts of SNM are reported through Nuclear Material Transaction Reports as required. As agreed to by the Licensee (ML23090A214), a TS under TS 6.0, Administrative Controls, was added to capture this procedure as part of the licensing basis (see TS 6.16.1 below). Therefore, the NRC staff finds that TMI-2 Solutions program meets the requirement of 10 CFR 74.15.

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TS 6.16.1

Establish written procedures to address applicable MC&A activities in 10 CFR 74.11, 74.13, 74.15, 74.19(a)(1) through (4), 74.19(b),and 74.19(c); however, for 74.19(a)(1) and 74.19(c), the scope is as described in Licensee response to RAI 14 and 16 (ML22276A024).

Recordkeeping (10 CFR 74.19(a))

The requirements at 10 CFR 74.19(a) states that a licensee is to keep records showing the receipt, inventory (including location and unique identity), acquisition, transfer, and disposal of all SNM in its possession regardless of its origin or method of acquisition. Each record relating to material control or material accounting must be maintained and retained for the period specified by the appropriate regulation or license condition. Each record of receipt, acquisition, or physical inventory of SNM must be retained as long as the licensee retains possession of the material and for 3 years following transfer or disposal of the material. Each record of transfer of SNM to other persons must be retained by the licensee who transferred the material until the Commission terminates the license authorizing the licensees possession of the material.

In the response to RAI 14 (ML22276A024), the Licensee states that the program description included in the response to RAI 10 describes the overall MC&A program and that requirements for records meeting 10 CFR 74.19(a) are included in the associated written procedures referred to in RAI 10. The Licensee states that there will be no receipts or acquisitions and therefore no records needed for those two activities. The Licensee states that records for inventory will be generated per written procedure, and that records for transfer or disposal will also be generated per written procedure. The Licensee states that record retention requirements are included in the TMI-2 records procedure which establishes retention requirements in 10 CFR 74.19(a)(2)-

(4). The Licensee affirms that records are generated per written procedure for inventory, transfer, and disposal of all SNM. The Licensee states that no receipts or acquisitions of SNM will occur and therefore the licensee will generate no records for those activities. The Licensee states that its records procedure established the records retention requirements in 10 CFR 74.19(a).

The NRC staff reviewed the Licensees description of MC&A records. Based on the review, the NRC staff has determined that the Licensees MC&A measures include adequate procedures to ensure MC&A records are completed and maintained. As agreed to by the Licensee (ML23090A214), a TS under TS 6.0, Administrative Controls, has been added to capture these procedures as part of the licensing basis (see TS 6.16.1 below). Therefore, the NRC staff finds that the Licensee meets the requirement of 10 CFR 74.19(a).

TS 6.16.1

Establish written procedures to address applicable MC&A activities in 10 CFR 74.11, 74.13, 74.15, 74.19(a)(1) through (4), 74.19(b),and 74.19(c); however, for 74.19(a)(1) and 74.19(c), the scope is as described in Licensee response to RAI 14 and RAI 16 (ML22276A024).

Written MC&A Procedures (10 CFR 74.19(b))

The requirements at 10 CFR 74.19(b) states that each licensee authorized to possess SNM in a quantity exceeding one effective kilogram shall establish, maintain, and follow written MC&A procedures sufficient to enable the licensee to account for the SNM in its possession under license.

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In the response to RAI 10 (ML22276A024 and ML23033A103), the Licensee provides the MC&A program description and the process description. The process described includes: SNM identification; SNM movement from in-situ to a package; preliminary accounting; final accounting; annual inventory; and annual reporting. For all steps in the process the Licensee states that the activities are performed in accordance with written procedures.

The NRC staff reviewed the Licensees description of MC&A procedures. Based on the review, the NRC staff determined that the Licensees MC&A measures include adequate procedures to ensure MC&A activities are performed to enable TMI -2 Solutions to account for the SNM in its possession under license. As agreed to by the Licensee (ML23090A214), a TS under TS 6.0, Administrative Controls, has been added to capture these procedures as part of the licensing basis (see TS 6.16.1 below). Therefore, the NRC staff finds that the Licensee meets the requirement of 10 CFR 74.19(b).

TS 6.16.1

Establish written procedures to address applicable MC&A activities in 10 CFR 74.11, 74.13, 74.15, 74.19(a)(1) through (4), 74.19(b),and 74.19(c); however, for 74.19(a)(1) and 74.19(c), the scope is as described in Licensee response to RAI 14 and 16 (ML22276A024).

Physical Inventory (10 CFR 74.19(c))

The requirements at 10 CFR 74.19(c) states that certain licensees who are authorized to possess SNM in a quantity greater than 350 grams of contained uranium-235, uranium-233, or plutonium, are to conduct a physical inventory of all SNM in its possession under license at intervals not to exceed 12 months. The results of these physical inventories shall be retained in records by the licensee until the Commission terminates the license authorizing the possession of the material.

In the response to RAI 16 (ML22276A024), the Licensee refers to an exemption granted by the NRC in 1985 (ML20138D392) for performing an annual physical inventory. At that time, annual physical inventories were required by 10 CFR 70.51(d). In a 2002 NRC rulemaking, 10 CFR 70.51(d) requirements were moved into Part 74 as 10 CFR 74.19(c) with little change. The Licensee states that the 1985 exemption from 10 CFR 70.51(d) is equivalent to an exemption today from 10 CFR 74.19(c) based on NRC staffs response (ML23069A158) to TMI-2 Solutions inquiry regarding applicability of exemptions. Notwithstanding its exemption, the Licensee states that upon packaging material in a container TMI-2 Solutions will perform inventory of the resulting containers per procedure. The Licensee also states the inventory will include SNM in any packages physically on site at the time of annual inventory. The Licensee further affirms that an annual inventory will be performed as required by 10 CFR 74.19(c).

The NRC staff reviewed the Licensees description of physical inventory. The NRC staff determined that the 1985 exemption from 10 CFR 70.51(d) is equivalent to an exemption today from 10 CFR 74.19(c). Based on the review, the NRC staff determined that the Licensees MC&A measures include adequate procedures to ensure physical inventories of its SNM are completed at the required frequency and the results are reported. As agreed to by the Licensee (ML23090A214), a TS under TS 6.0, Administrative Controls, has been added to capture these procedures as part of the licensing basis (see TS 6.16.1 below). Therefore, the NRC staff finds that the Licensee meets the requirement of 10 CFR 74.19(c).

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TS 6.16.1

Establish written procedures to address applicable MC&A activities in 10 CFR 74.11, 74.13, 74.15, 74.19(a)(1) through (4), 74.19(b),and 74.19(c); however, for 74.19(a)(1) and 74.19(c), the scope is as described in Licensee response to RAI 14 and RAI 16 (ML22276A024).

MC&A Conclusion

Based on the review of the TMI-2 Solutions LAR, submitted supplements, responses to RAIs and clarifications, and the addition of TS 6.16, the NRC staff finds that the Licensees MC&A practices as described satisfy the applicable requirements at 10 CFR 74.11, 74.13, 74.15, and 74.19(a)(1) through (4), 74.19(b),and 74.19(c); however, for 74.19(a)(1) and 74.19(c), the scope is as described in Licensee response to RAI 14 and RAI 16 (ML22276A024) during the license term. Therefore, the NRC staff finds that the Licensees proposed MC&A programs and practices are acceptable.

3.2.6 TS Section 3/4.3, Crane Operations

TS Section 3/4.3, Crane Operations, includes LC for PDMS 3.3.1 which states that loads in excess of 50,000 lbs. shall be prohibited from travel over the RV unless a docketed Safety Evaluation for the activity is approved by the NRC. In its LAR, the Licensee proposed to delete TS Section 3/4.3, Crane Operations, because TS 3/4.3 does not satisfy any of the four requirements established in 10 CFR 50.36(c)(2)(ii). The NRC staffs evaluation of the satisfaction of the four requirements are discussed below in further detail.

Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Licensees Evaluation

The Licensee indicates that TMl-2 does not have a reactor coolant pressure boundary.

The Licensee concluded that Criterion 1 is not applicable to TMl-2.

NRC Staff Evaluation

The NRC staff evaluated the Licensees response and finds that since TMI -2 no longer has a reactor coolant pressure boundary, TS Section 3/4.3 no longer meets Criterion 1.

Criterion 2: 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.

Licensees Evaluation

The Licensee indicated that TMI-2 has removed 99 percent of its fuel, the facility is no longer licensed for power operations, and that currently there are no DBAs or transients associated with TMl-2. The Licensee concluded that Criterion 2 is accordingly not applicable to TMl-2.

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NRC Staff Evaluation

The NRC staff evaluated the information provided in the Licensees submittal and agrees with the Licensees assessment that TMI-2 is defueled, has a POL, and there are no DBAs or transients associated with the facility. The staff noted that the Licensee performed a containment fire analysis, and it does not credit any TMI-2 crane in its mitigation. Therefore, the NRC staff finds that TS Section 3/4.3 no longer meets Criterion 2.

Criterion 3: A structure, system, or component 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.

Licensees Evaluation

The Licensee indicated that there are no DBAs or transients associated with TMl-2 and the cranes do not provide a required function to mitigate the effect of unanticipated occurrences. The Licensee concluded that Criterion 3 is not applicable to TMl-2.

NRC Staff Evaluation

The NRC staff evaluated the information provided in the Licensees submittal and finds that TMI-2s license is a POL, the TMI-2 reactor is defueled with 99 percent of its fuel has been removed, and there are no DBAs or transients associated with the site. The staff noted that the applicant performed a containment fire analysis, and it does not credit TMI-2 crane in its mitigation analysis. Therefore, there are no SSCs in the crane system that represent a challenge to the integrity of a fission product barrier, and the NRC staff finds that TS Section 3/4.3 no longer meets Criterion 3.

Criterion 4: A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

Licensees Evaluation

The Licensee indicated that TMI-2 has removed 99 percent of its fuel and that currently there are no DBAs or transients associated with TMl-2. The Licensee states that it performed a calculation TMl2-EN-RPT-0001, "Determination of the Safe Fuel Mass Limit for Decommissioning TMl-2," and submitted it as Attachment 5 to the LAR, as updated per Attachment 1 to TMI-2 Solutions submittal dated April 7, 2022 (ML22101A077), that demonstrates that the remaining core debris material cannot be configured into an arrangement whereby a criticality event is credible. The Licensee accordingly concluded that Criterion 4 is not applicable to TMl-2.

NRC Staff Evaluation

The NRC staff evaluated the information provided in the Licensees submittal and finds that TMI-2 is no longer licensed to operate, there are no DBAs or transients associated with the site, and that even in the case a load is dropped the remaining fuel mass is insufficient to achieve criticality. The NRC staff finds that the there are no SSC related to TS Section 3/4.3 that are significant to public health and safety, therefore, TS 3/4. 3 no longer meets Criterion 4.

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The NRC staff reviewed the four criterion in 10 CFR 50.36(c)(2)(ii) and determined that TS 3/4.3 do not satisfy any of the requirements and therefore, can be removed from TS. The NRC staff concludes that these changes are administrative in nature.

3.2.7 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 on 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 intake values. The TMl-2 sealed sources are no longer managed by the TMI-1 licensee, Constellation Generation Energy, LLC. The sealed sources at TMI-2 are maintained by TMI -2 Solutions under a program compliant with the requirements of 10 CFR 70.39(c) (ML23033A103).

The Licensee stated that TS 3/4.4 does not satisfy any of the four requirements in 10 CFR 50.36(c)(2)(ii) for equipment and parameters to be included in the TS.

Licensees Evaluation:

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.

NRC Staff Evaluation

The NRC agrees that TMI-2 does not have a reactor coolant pressure boundary and therefore the requirements of Criterion 1 of 50.36(c)(2)(ii)(A) are not applicable. Therefore, the NRC finds it acceptable to remove TS 3/4.4.

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."

Licensees Evaluation

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.

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NRC Staff Evaluation

The NRC agrees that TMl-2 is no longer licensed to operate (ML19208A147), therefore the requirements of Criterion 2 of 10 CFR 50.36( c)(2)(ii)(B) are not applicable. Therefore, the NRC finds it acceptable to remove TS 3/4.4.

Criterion 3 - The requirements of Criterion 3 of 10 CFR 50.36(c)(2)(ii)(C) states that TS 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."

Licensees Evaluation

TMI-2 Solutions states in its LAR that sealed sources do not provide a function required to mitigate the effect of unanticipated occurrences such as the fire in containment as described on the update to the containment fire scenario fire analysis (ML23033A103). Thus, the requirements of 10 CFR 50.36(c)(2)(ii)(C) do not apply.

NRC Staff Evaluation

The NRC staff concludes that sealed sources do not provide a function required to mitigate the effects of unanticipated occurrences and thus the requirements of 10 CFR 50.36(c)(2)(ii)(C) do not apply. Therefore, the NRC finds it acceptable to remove TS 3/4.4.

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."

Licensees evaluation

TMI-2 Solutions states that the intent of this criterion is for risk insights and operating experience be factored into the establishment of TS LCOs. Further, TMI-2 Solutions states there are no TS associated with Phase 1 b or Phase 2, hence there are no limiting conditions for PDMS. Also, TMI-2 Solutions states that it has a procedure that provides a methodology for gathering OPEX/LL information (Operating Experience/Lesson Learned) from various sources for systematic review, evaluation, and use on the TMl-2 project. TMI-2 Solutions contends that since the containment fire analysis does not credit containment closure and assumes a release to the environment, which is bounded by the results of the updated containment fire scenario fire analysis (ML23033A103), the requirements of Criterion 4 of 10 CFR 50.36 (c)(2)(ii)(D) are not applicable. Therefore,TMI-2 Solutions concludes that the requirements of TS 3/4.4 be deleted from the TS and controlled under a program which meets the requirements of 10 CFR 70.39(c) and will be subject to 10 CFR 50.59 change evaluation process that ensures adequate regulatory controls are in place.

NRC Staff Evaluation

The NRC staff reviewed the Licensees proposed relocation of TS 3/4.4 and agrees that TS 3/4.4 does not satisfy any of the requirements in 10 CFR 50.36(c)(2)(ii) and therefore, can be removed from TS. The NRC staff finds that relocation of these requirements to the DSAR is acceptable to be controlled in a place other than TS under a program which meets the

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requirements of 10 CFR 70.39(c) and will be subject to 10 CFR 50.59 change evaluation process that ensures adequate regulatory controls are in place for sealed sources which do not meet the requirements of 10 CFR 50.36(c)(2)(ii).

The NRC staff agrees that relocating TS 3/4.4 to the DSAR is appropriate since the DSAR reflects the permanently shutdown and defueled condition and changes are controlled by the 10 CFR 50.59 process. Further, since the TMl-2 sealed sources are maintained at TMl-1 and managed by the TMI-1 licensee, Constellation Generation Energy, LLC, under a program compliant with the requirements of 10 CFR 70.39(c), the TMI-2 sealed sources will remain under regulatory control. The NRC concludes that removal of TS 3/4.4 and relocation to the DSAR is acceptable. The NRC concludes that deleting TS 3/4.4 is administrative in nature.

3.2.8 TS Section 5.0, Design Features

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 in the LAR, TMI-2 Solutions contents there is justification for deleting containment related Limiting Conditions for PDMS and Surveillance Requirements because the worst-case accident at TMI -2 will not have a significant effect on safety. Additionally, it states that this same information can be found in the PDMS SAR.

TMI-2 Solutions proposes to remove TS 3/4.0 Limiting Conditions and Surveillance Requirements since they are no longer needed to support transition from PDMS to DECON.

The NRC concludes that s ince the licensee is removing TS Section 3/4.1 through 3/4.4, and the NRC finds it acceptable, there is no need to define limiting conditions or associated Surveillance Requirements for this section. See Section 3.7.3 above. Therefore, the bases associated with Sections 5.0 is removed. Therefore, the NRC staff finds TS Section 5.0 unnecessary and concludes it acceptable to remove them in its entirety. The NRC staff concludes that deleting TS 5.0 is administrative in nature.

3.2.9 TS Section 5.1 Containment

TS Section 5.1 "Containment, Configuration" identifies principal design parameters associated with a robust containment structure designed to accommodate pressurization.

Licensees Evaluation:

In its February 19, 2021, LAR, the Licensee proposed to delete TS Section 5.1, because the TMI-2 containment structure no longer satisfies any of the four requirements established in 10 CFR 50.36(c)(2)(ii).

NRC Staff Evaluation

TMI-2 Solutions proposes to remove TS 3/4.0 Limiting Conditions and Surveillance Requirements (SRs) since they are no longer needed to support transition from PDMS to DECON.

The NRC concludes that since the licensee is removing TS Sections 3/4.1 through 3/4.4, and the NRC finds it acceptable, there is no need to define limiting conditions or associated SR for

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this section. See Section 3.2.3 above. Therefore, the bases associated with Sections 5.1 is removed. Therefore, the NRC staff finds TS Section 5.1 unnecessary and concludes it acceptable to remove them in its entirety.

Further, the NRC staffs evaluation of the satisfaction of the four requirements are discussed below in further detail.

Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Licensees General Evaluation of Criterion 1

TMI-2 Solutions indicates that TMl-2 does not have a reactor coolant pressure boundary.

The Licensee concluded that Criterion 1 is not applicable to TMl-2.

Staff evaluation

The NRC staff evaluated the Licensees response and finds that since TMI -2 no longer has a reactor coolant pressure boundary, TS Section 5.1 no longer meets Criterion 1.

Criterion 2: 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.

Licensees General Evaluation of Criterion 2

The licensee indicated that TMI-2 has removed 99 percent of its fuel and that currently there are no DBAs or transients associated with TMl-2. The licensee concluded that Criterion 2 is not applicable to TMl-2.

Staff evaluation

The NRC staff evaluated the information provided in the Licensees submittal and finds that TMI-2 is no longer licensed to operate, 99 percent of its fuel has been removed and that there are no DBAs or transients associated with the site. The NRC staff noted that the applicant performed a containment fire analysis, and it does not credit containment closure or isolation in its mitigation. Therefore, the NRC staff finds that TS Section 5.1 no longer meets Criterion 2.

Criterion 3: A structure, system, or component 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.

Licensees Evaluation of Criterion 3

The Licensee indicated that there are no DBAs or transients associated with TMl-2. The Licensee concluded that Criterion 3 is not applicable to TMl-2.

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Staff evaluation

The NRC staff evaluated the information provided in the Licensees submittal and finds that TMI-2 is no longer licensed to operate, 99 percent of its fuel has been removed and that there are no DBAs or transients associated with the site. The NRC staff noted that the applicant performed a containment fire analysis, and it does not credit containment closure or isolation in its mitigation. Therefore, the NRC staff finds that TS Section 5.1 no longer meets Criterion 3.

Criterion 4: A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

Licensees Evaluation

The Licensee indicated that TMI-2 has removed 99 percent of its fuel and that there are no DBAs or transients associated with TMl-2. The Licensee concluded that Criterion 4 is not applicable to TMl-2.

NRC Staff evaluation

The NRC staff evaluated the information provided in the LAR and finds that TMI-2 has a POL, is permanently defueled and that there are no DBAs or transients associated with the site. This has reduced the risk of an accident capable of significantly impacting the public health and safety. The staff also noted that the applicant performed a containment fire analysis (which is not considered a DBA), and the mitigation of this event does not credit containment closure or isolation. The consequences of the containment f ire are bounded by the results of the updated containment fire scenario fire analysis (ML23033A103), and remain below regulatory limits. The NRC staff has not identified any operational experience that would require crediting TMI-2 containment systems for the protection of public health and safety. The NRC staff finds that the containment systems are not significant for the protection of public health and safety, therefore, TS Section 5.1 no longer meets Criterion 4.

The NRC staff evaluated the criterion identifies in 10 CFR 50.36(c)(2)(ii)(A)-(D) and determined that TS 5.1 no longer meets the criterion for inclusion in the TS and therefore it is acceptable to delete these TS.

Therefore, the NRC staff concludes that deleting TS Section 5.1 is administrative in nature.

3.2.10 TS Section 6.0, Administrative Controls

TS 6.1, RESPONSIBILITY

TS 6.1.1 describes the responsibility of the project director. The Licensee proposes to change the term unit to facility and relocate this TS to the DQAP.

The NRC staff finds the proposed changes acceptable because relocating this TS to the DQAP is an administrative change and does not change any qualifications, requirements, or responsibilities of the position and is consistent with NRC Administrative Letter 95-06 (ML20101P963). The NRC staff also finds that the revision to replace unit with facility more

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accurately reflects the permanently shutdown and defueled condition and is therefore acceptable. The NRC staff concludes that the changes are administrative in nature.

TS 6.2, ORGANIZATION

The Licensee proposes to delete TS 6.2.1 pertaining to Facility Organization since this information will be conveyed in the DQAP and there is no need to refer to the PDMS SAR.

The Licensee proposes to revise TS 6.2.2 to reflect that the information pertaining to Organization will be conveyed in the DQAP and there is no need to refer to the PDMS SAR. In addition, the Licensee proposes to replace the term unit with facility in the title of TS 6.2.2 more accurately reflect the status of the permanently defueled and non-operational TMI-2.

The NRC staff finds the proposed changes acceptable because deleting TS 6.2.1 and relocating TS 6.2.2 to the DQAP is consistent with NRC Administrative Letter 95-06. The NRC staff also finds that the revision to replace unit with facility more accurately reflects the permanently shutdown and defueled condition and is therefore acceptable. The NRC staff concludes that the changes are administrative in nature.

TS 6.3, UNIT STAFF QUALIFICATIONS

TS 6.3 is an administrative control that addresses Unit Staff Qualifications. TS 6.3.1 and TS 6.3.2 are subsets of TS 6.3 that discuss staff qualifications. Both TS 6.3.1 and TS 6.3.2 uses the term unit and TMI-2 solutions proposes to change the term :unit to facility in both places to reflect that the reactor is no longer authorized to operate. Also, TMI-2 Solutions proposes to delete and relocate TS 6.3.1 and TS 6.3.2 to the DQAP. TMI-2 Solutions states that NRC Administrative Letter 95-06 allows for the relocation of TS administrative controls to a quality assurance program. Further in the LAR, TMI-2 states that Administrative Letter 95-06 explains that a relocation is acceptable to the NRC because of the controls imposed by 10 CFR Part 50 Appendix B, the requirement for Part 50 licensee to have an NRC approved quality assurance program, and the provision in the regulations for a quality assurance program change control process in 10 CFR 50.54(a). TMI-2 Solutions contends that it meets these criteria because it must meet 10 CFR 50 requirements. TMI-2 Solutions also proposes that after TS 6.3 be deleted and relocated into the DQAP. TMI-2 Solutions states that any future changes will be controlled in accordance with 10 CFR 50.54(a).

The NRC staff finds the proposed changes acceptable because deleting 6.3.2 and relocating TS 6.3.1 to the DQAP is consistent with NRC Administrative Letter 95-06. The NRC staff also finds that the revision to replace unit with facility more accurately reflects the permanent shutdown and defueled condition of TMI-2 and is therefore acceptable. The NRC staff concludes that the changes are administrative in nature.

TS 6.6 REPORTABLE EVENT ACTION

TS 6.6 specifies that Licensee Event Reports (LER) shall be submitted pursuant to the requirements of 10 CFR 50.73. The Licensee proposes to delete TS Section 6.6 in its entirety because as the actions of this section are required by regulation, it is not necessary to restate these requirements in a TS.

The NRC staff reviewed the proposed deletion and agreed that TS 6.6 is not necessary because it is redundant to the requirements of 10 CFR 50.73. Therefore, the NRC staff finds that the proposed deletion of this TS is acceptable. The NRC staff concludes that the changes are administrative in

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nature.

TS 6.7 PROCEDURES AND PROGRAMS

TMI-2 Solutions proposes to delete TS 6.7, Written Procedures and Programs because they are addressed in the ODCM. The Licensee also proposes to revise TS 6.7.1 to delete references to PDMS. In its January 27, 2023, LAR supplement (ML23033A103), the Licensee proposes to correct an inadvertent deletion of language in TS 6.7.1 (ML22013A177) from its previous submittal (ML21057A047) and proposes to keep the phrase as described in the SAR.

The Licensee proposes to relocate TS 6.7.2 to the DQAP with no text changes, thus deleting it from the TS. The Licensee proposes to revise TS 6.7.3.b, 6.7.4.a.4, 6.7.4.a.8, and TS 6.7.4.a.9 to replace unit with facility to better reflect the nature of TMI-2s status. The Licensee proposes to revise TSs 6.7.4.a, 6.7.4.a.4, 6.7.4.a.9, and 6.7.4.a.10 to delete MEMBERS OF THE PUBLIC and replace with lower case wording. The Licensee proposes to revise TS 6.7.4.a.2 to delete UNRESTRICTED AREAS and replace with lower case wording. Finally, the Licensee proposes to delete TS 6.7.4.b.2 and 6.7.4.b.3, which are both currently controlled by the ODCM and are therefore redundant TSs.

The NRC staff concludes that the changes are administrative in nature. The NRC staff finds that relocation of the information to the DSAR is acceptable.

The NRC staff finds the proposed changes acceptable because relocating TS 6.7.1, 6.7.2, 6.7.3, and 6.7.4 (with the exceptions noted in the LAR) to the DQAP and relocating TS 6.7.4.b.2 and TS 6.7.4.b.3 to the ODCM is consistent with NRC Administrative Letter 95-06. Therefore, the NRC staff finds the proposed revision to TS 6.7.1 to delete reference to the PDMS condition acceptable and it supports transition from PDMS to DECON. The NRC staff also finds that the revisions to replace unit with facility more accurately reflect the permanently shutdown and defueled condition and is therefore acceptable. The NRC staff also finds the revisions to the terms members of the public and unrestricted areas acceptable because those are defined terms in 10 CFR 20.1003 and therefore, need not be defined in the TS. The NRC staff concludes that the changes are administrative in nature.

TS 6.8 REPORTING REQUIREMENTS

The Licensee proposes to delete portions of TS 6.8, Reporting Requirements, and relocate it to the DQAP. However, the Licensee will maintain TS 6.8.1.2 in its POL TS. In addition, the Licensee proposes the following revisions to the TS. The licensee proposes to delete duplicative language that refers to reporting requirements from what appears in NRC regulations in TS 6.8.1. The Licensee proposes to replace unit with facility in TS 6.8.1.1, TS 6.8.1.2, and TS 6.8.1.4. The Licensee proposes to revise TS 6.8.1.4.a to reflect approved PDMS SAR changes that are required to be presented in the DSAR update as required by 10 CFR 50.71(e).

The NRC staff finds the proposed changes acceptable because relocating portions of TS Sections 6.8.1, 6.8.2, and 6.8.3 to the DQAP is consistent with NRC Administrative Letter 95-06.

The NRC staff also finds that the Licensees deletion of language in 6.8.1 is acceptable as that language is unnecessary as the actions in that section are required by NRC regulation. The NRC staff also finds that the revisions to replace unit with facility more accurately reflect the permanently shutdown and defueled condition and is therefore acceptable. The NRC staff also finds the revisions to TS 6.8.1.4.an acceptable because they accurately reflect the approved PDMS SAR and FSAR update. The NRC staff concludes that the changes are administrative in nature. The NRC staff finds that relocation of the information to the FSAR, as stated above (to

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DSAR), is acceptable to the NRC staff.

TS 6.9 RECORD RETENTION

The Licensee proposes to delete TS 6.9, Records Retention and relocate it to the DQAP. In addition, the Licensee proposes to revise language in TS 6.9.2 to reflect that the TMI-2 license is a possession only license. The Licensee proposes to revise the term reportable events in TS 6.9.2.c to reflect that the term is not defined in the TS (i.e., it will no longer be shown in capital letters) since the definition is in the applicable regulations of 10 CFR 50.72 and 10 CFR 50.73. The Licensee proposes to modify TS 6.9.2.e to clarify that records of changes to procedures associated with the Recovery Technical Specifications will be retained, and that records of changes to procedures and programs as specified in PDMS TS 6.7.1 are retained.

The Licensee proposes to define acronyms Technical Evaluation Report (TER) and System Description (SD) in TS 6.9.2.i for ease of reading. The Licensee proposes to replace unit with facility in TS 6.9.2.m and 6.9.2.s to better reflect the current status of TMI-2. The Licensee proposes to clarify in TS 6.9.2.q that the records associated with the Generation Review Commission (GRC), and Independent Onsite Safety Review Group (IOSRG), as applicable to TMI-2, will be retained.

The NRC staff finds the proposed changes acceptable because the requirements for record retention can be adequately addressed by the DQAP (10 CFR 50, Appendix B, Criterion XVII) and because provisions relating to record keeping support the assurance of safe operation of a facility in a permanently defueled condition. The NRC staff notes that 10 CFR 20, Subpart L and 10 CFR 50.71 require retention of certain records related to the facility. Since record retention will be maintained by the DQAP, consistent with NRC Administrative Letter 95-06, the NRC staff finds this change acceptable. The NRC staff also finds that the Licensees revision to state that TMI-2 is not licensed to operate but only to possess fuel is acceptable since it accurately reflects the facilitys permanently defueled condition. The NRC staff also finds that the Licensees revision to reportable events acceptable since this is defined in 10 CFR 50.72 and 10 CFR 50.73. The NRC staff finds the Licensees revision to 6.9.2.e acceptable since it appropriately retains records of changes to procedures and programs. The NRC staff finds the Licensees revision to define TER and SD in TS 6.9.2.i acceptable as the terms will be used in the TS more than once. The NRC staff also finds that the revisions to replace unit with facility more accurately reflect the permanently shutdown and defueled condition and is therefore acceptable. The NRC staff also finds the Licensees clarification regarding retaining the records associated with the GRC and IOSRG acceptable as they are editorial. The NRC staff concludes that the changes are administrative in nature.

TS 6.10 RADIATION PROTECTION PROGRAM

The Licensee proposes to delete TS Section 6.10 related to its RPP from its TS and relocate them to the DQAP. Since the RPP requirement will be maintained by the DQAP, the NRC staff finds the change acceptable and note that it is consistent with NRC Administrative Letter 95-06.

The NRC staff concludes that the change is administrative in nature.

TS 6.11 HIGH RADIATION AREA

The Licensee proposes to change TS 6.11 to incorporate an alternate High Radiation Area (HRA) Control for TMI-2, which meets the requirements of 10 CFR 20.1601. The a lternate control proposed would provide greater operational efficiency by reducing the number of areas

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within the facility required to be maintained as locked to prevent unauthorized or unintended access.

As stated in 10 CFR 20.1601(a), licensees are required to institute strict controls for access to areas where an individual could receive greater than 100 mrem in one hour. A list of optional controls for HRAs is provided in 10 CFR 20.1601(a) and (b). For large complex facilities, such as a nuclear power plant, the most practical of these options is to maintain the entrances to HRAs locked except during periods of authorized personnel access. However, there are circumstances where the physical arrangement and access requirements of HRAs make this control method impractical. In cases where the controls provided in 10 CFR 20.1601(a) and (b) unnecessarily restrict plant operations,10 CFR 20.1601(c) provides for the licensee to propose alternative controls for access to HRAs.

Section 2.4 of Regulatory Guide 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants (ML061350096), describes an acceptable alternative to maintaining all HRAs locked. Under this alternative control scheme, areas where individuals could receive doses greater than 100 mrem, but less than or equal to 1000 mrem, in an hour are barricaded and conspicuously posted in lieu of being maintained locked. Areas where doses in excess of 1000 mrem, but less than 500 rad, could be received in one hour are maintained locked, except for special cases where individual areas cannot reasonably be enclosed, a barricade with a flashing-light warning device is provided. As a compensatory measure, access to a HRA may be controlled by the issuance of a radiation work permit (RWP) to ensure that individuals are appraised of the known radiological conditions and protective actions required for accessing the area. Additional radiation monitoring is also required for individuals entering a HRA to provide for possible unanticipated exposure situations.

The Licensees proposed change would:

• Invoke the exception of 10 CFR 20.1601(c) for TMI-2 for alternate requirements in lieu of the control device and alarm signal requirements of 10 CFR 20.1601(a) and (b)

• Remove the differentiating HRA limit of 1000 mrem/hour (1 rem/hour) because the HRA requirements are now also applicable to areas > 1000 mrem/hour

• Provide for alternate entrance requirements to HRAs consistent with Regulatory Guide 8.38, Section C, Paragraph 2.4, including issuance of an RWP, or equivalent, or by being qualified in, or escorted by individuals qualified in, radiation protection procedures regarding work in high radiation areas

• Provide an additional alternate HRA entrance requirement to include use of appropriate radiation dose rate monitoring device(s)

• Continue requirements for a locked or continuously guarded door, or gate, or equivalent to prevent entry to HRAs in which an individual could be exposed to a dose rate > 1000 mrem/hour.

• Change the key control from administrative control of the Radiological Controls Supervisor to administrative control in accordance with a program approved by the radiation protection manager

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• Make the TMI-2 TS for control of HRAs consistent with that previously approved for TMI -

1

The TMI-2 TS will include all the above alternative methods of control, except for the flashing warning light alternative. TMI-2 Solutions states in its LAR that a flashing light shall be activated as a warning device.

Licensees Evaluation

TMI-2 Solutions indicates in its proposed changes that it proposes to use 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, 10 CFR 20.1601 is alternative criteria for High Radiation Areas.

TMI-2 Solutions states in its LAR that the TS changes proposed are consistent with those proposed by Exelon for TMl-1 (Reference 27 (ML19178A304)) and approved by the NRC (Reference 28)(ML20134H940) for TMl-1. TMI-2 Solutions explains that language in the LAR is consistent with and bounds the control device and alarm signal exceptions. Further, TMI-2 Solutions states that the key control will be in accordance with a program approved by the radiation protection manager.

TMI-2 Solutions proposes to develop procedures associated with Phase 1b will be developed to retrieve the remaining core debris material and decontaminate high radiation areas. TMI-2 Solutions proposes to also develop Phase 2 procedures for decommissioning of a reactor facility. TMI-2 Solutions explains in the LAR that the Radiation Protection Program 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. TMI-2 Solutions concludes in the LAR that the containment isolation TS Limiting Conditions for PDMS are not required, and do not apply in Phase 1 b and Phase 2.

Further, TMI-2 Solutions, contends that the TS change reorganizes the TS into a clear sequence and separates the TS into sections so that the subject matter is more easily recognized and understood. Also, TMI-2 Solutions states that the TS wording is revised to improve worker efficiency, increase awareness, clarify requirements, and enhance readability.

Also, TMI-2 Solutions states that proposed TS wording is consistent with existing standard TS (STS) format (Reference 26) of using two main sections (one for high radiation areas and one for locked high radiation areas), with edits to remove redundancy and to improve clarity and readability.

NRC Staff Evaluation

The proposed changes are only related to the control of access to high radiation areas to minimize dose to plant personnel. The NRC staff reviewed the proposed changes against Reg Guide 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants," published in May 2006 and found the changes were consistent with this guidance. The NRC concludes that the proposed changes are intended to provide clarity and/or flexibility with respect to the administration and programmatic controls while retaining adequate margin of safety for minimizing dose to site personnel consistent with the requirements of 10 CFR 20, "Standards for Protection Against Radiation," and the guidance of RG 8.38. Since there are no associated physical plant changes, the ability of the plant to respond to and mitigate accidents is unchanged by the proposed changes. Therefore, the NRC staff finds the proposed changes are

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acceptable. The changes would allow for alternative methods for the control of access to High Radiation Areas per 10 CFR Part 20.

TS 6.12 ODCM

The Licensee proposes to delete TS Section 6.12 related to the ODCM from its TS and relocate it to the DQAP. Since the ODCM requirement will be maintained by the DQAP, consistent with NRC Administrative Letter 95-06, the NRC staff finds this change acceptable. The NRC staff concludes that the changes are administrative in nature.

TS 6.13 EXCEPTIONAL OCCURRENCES

The Licensee proposes to delete TS Section 6.13, exceptional occurrences, from its TS and relocate them to the DQAP. Since the exceptional occurrences requirement will be maintained by the DQAP, consistent with NRC Administrative Letter 95-06, the NRC staff finds this change acceptable. The NRC staff concludes that the changes are administrative in nature.

4.0 PUBLIC COMMENT

On August 22, 2022, the NRC published a Federal Register notice (87 FR 51454, 51460) describing the LAR and describing the NRC staffs proposed determination that the LAR involved no significant hazards consideration and informing the public of the opportunity to submit comments by September 21, 2022.

There were no comments received in response to the Federal Register notice.

5.0 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

The Commission issued a proposed finding that the amendment involves no significant hazards consideration, published in the Federal Register on August 22, 2022 (87 FR 51454), and there have been no public comments on this preliminary finding. TMI-2 Solutions later supplemented its application and modified its No Significant Hazard Determination (NSHD) (ML23033A103).

The NRC staff did not reissue the Federal Register notice because the modified NSHD was bounded by the previous NSHD in scope and the NRC staff found the change in accident analysis were bounded by the first proposed NSHD that the NRC published in the Federal Register on August 22, 2022. There were no public comments on this preliminary finding.

The NRC staff determined that there is no significant hazards consideration related to this LAR.

Accordingly, the NRC staff finds that, under 10 CFR 50.92 Issuance of amendment, the operation of TMI-2 in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

As part of its NSHC determination, the NRC staff considered whether there was any potential of radiological consequences related to this LAR and whether criticality was credible based on the updated SFML. Each of these considerations is discussed in more detail below.

5.1 Radiological Consequences (Accident Analysis)

The NRC staff reviewed TMI-2s accident analysis evaluation to ensure that there would be no

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impact to the TS. In accordance with 10 CFR 50.36(c)(6), TS involving safety limits, limiting safety system settings, and limiting control system settings; limiting conditions for operation; surveillance requirements; design features; and administrative controls will be developed on a case-by-case basis for reactors in decommissioning. The NRC staff evaluated accident scenarios to ensure that the proposed changes are safe and do not impact the NRC Staffs NSHD.

The NRC performed a detailed review of the materials submitted by TMI -2 Solutions including examination of references, background materials, and independent verification of a sufficient number of the Licensees calculations in the LAR.

The Licensee considered both a 1 mrem CEDE limit and 2 times the organ dose rate from the ODCM as constraints for developing operational activity limits. Because the ODCM limit is applied to a 60-minute time period and the limiting organ was the bone, the equivalent Committed Dose Equivalent (CDE) limit to the bone was 3000 mrem / 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> or 0.342 mrem over a 60-minute period. This limit is far more restrictive than the 1 mrem CEDE. For example, for Am-241 (on a per unit of radioactivity basis) the CDEbone is a factor of 18 more restrictive (a factor of 53 overall). The ODCM limits that are cited are for effluents that apply to expected or normal conditions. The requirements at 10 CFR 50.36a(a) are provided To keep releases of radioactive materials to unrestricted areas during normal conditions, including expected occurrences, as low as is reasonably achievable(emphasis added). A fire of sufficient magnitude that releases radioactivity offsite is an unexpected release (accident) that is not anticipated to occur. The EPA P rotective Action Guide (PAG) provides a limit for an offsite radiological dose to a member of the public in the range of 1 to 5 rem from a radiological incident or accident.

The EPA PAG guidelines are applicable to any radiological incident or accident. The setting of emergency action levels is designed to ensure that proper responses can be taken to ensure EPA PAG guidelines will not be exceeded. The exceedance of EPA PAG guidelines depends on the radiological source term, the form of the radiological source term, and the mechanisms and processes of release. Without sufficient source term, a release may trigger a NOUE, but sufficient radioactivity may not be present to trigger higher emergency action levels. The defueled TMI-2 reactor contains fuel fragments and other materials (termed debris material) but the quantity is roughly 1 percent of the initial fuel. The debris material is widely disbursed throughout the reactor systems and building. TMI -2 does not have any spent fuel in a spent fuel pool, which is the primary consideration for NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors. Given the unique conditions of TMI -2, use of a 1 mrem CEDE is sufficiently protective as a NOUE emergency action level, though the Licensee may elect to use a more restrictive limit to provide an additional safety margin.

NRC staff evaluated the scope of the Licensees evaluation of potential accidents that may produce an offsite radiological impact and found it to be sufficient. The most likely accidents are a drop of a container or other form of mechanical disturbance during decommissioning activities.

Those scenarios involve disturbance of small amounts of radioactivity and are bound by updated containment fire scenario fire assessment (ML23033A103). The next most likely type of accident, though not expected, is a fire. Fire is arguably one of the largest risks at a nuclear facility (DOE, 1994). Fire risk is a product of the likelihood of a fire occurring and the consequences if a fire were to occur. Fires have occurred at nuclear reactors undergoing decommissioning though they have been very minor i n impact. 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

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than during power operations. The Licensee evaluated a diverse set of potential fires and derived operational limits to ensure that potential offsite releases, if they were to occur, would be meet established requirements. The scope of the fire analyses included consideration of what could catch fire and where it could catch fire. The licensee also evaluated other scenarios that were previously considered to determine that they were still valid as well as new scenarios introduced by the NRC as potentially applicable to the DECON phase. Each of these scenarios is discussed below.

The Licensees initial fire evaluation, based on past precedent, was an outdoor fire involving spent ion exchange resins. This scenario was appropriate if the resins used were organic or otherwise combustible. For the processing of contaminated water, the Licensee states it will use a zeolite-based resin which is not combustible, thereby eliminating the scenario. The NRC staff agrees with the Licensees conclusion.

Next, the Licensee used previously examined fire scenarios to demonstrate that potential accidents during decommissioning would not result in releases to a member of the public at the site boundary that would exceed the lower range of the EPA PAG of 1 rem. The Licensee assessed a fire in the Reactor Building basement and estimated that the resultant inhalation dose would be a maximum of 97 mrem with the HEPA filters not functioning for an elevated release. The limit the Licensee was attempting to meet was 100 mrem TEDE. Because the result was close to the desired limit, the NRC staff had to assess if the conservatisms introduced accounted for the uncertainties in the calculations. There were a variety of uncertainties associated with the analysis that NRC staff issued a request for additional information (ML22210A080). These uncertainties included:

• Values for the airborne release fraction (ARF)

• Scope of the fire scenarios

• Basis for the offsite dose calculations

• Radiological inventory

• Buildup of radiolytic gases (hydrogen)

• Dust explosions and exothermic reaction hazards

• A fire involving cork seams

The Licensee provided additional information in response to the request (ML23033A103).

In the analyses of potential fires involving DAW or a fire in the reactor building, the Licensee used a value of 1.5 x 10-4 for the ARF. It is appropriate to assign an ARF commensurate with the material that is being evaluated. Much of the material remaining in the TMI-2 reactor systems and buildings that are contaminated is not readily combustible or a source of fuel for a fire (e.g., metals, concrete, and sediment material in the Reactor Building basement). However, some material that is potentially combustible remains (e.g., plastics, wood). In an RAI, the NRC staff summarized measurements of ARFs for different materials (ML22210A080). ARFs can range over many orders of magnitude. In response, TMI-2 Solutions indicated that the radioactive material in the combustible waste present is Dry Active Waste (DAW) (e.g., waste bags, and materials, disposable protective clothing) that is contaminated with removable contamination. The removable contamination is generally in the form of non-combustible dust particles or powders that get consumed or entrained in the fire. The ARF used was from Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Power Station, NUREG/CR-0130, Vol 2 (1978), "release fraction is assumed to be equal to the release fraction from a contaminated waste fire, or 1.5 x 10-4" (NUREG/CR-0130, 1978). The Licensee

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summarized values applicable to the types of materials present in TMI-2, stating that representative ARF values ranged from 6 x 10-5 to 1 x 10-4, and therefore the value used was appropriate. The ARF values for certain materials, such as uncontained plastics, can be considerably higher. When assessing potential inhalation doses, the ARF is combined with a respirable fraction (RF). The product of the two factors (ARF and RF) is generally lower than the ARF alone by an order of magnitude. The Licensee used ARF in place of the product of the two factors because it assumed all the material that is released to the air is respirable. The staff concludes that the Licensees selection of 1.5 x 10-4 for the ARF/RF product for DAW with removable contamination is within the range of values observed. While this value is not large enough for uncontained plastics and wood, as further discussed in this section, other conservatisms introduced by the Licensee mitigate the impact of the selected ARF.

The Licensee derived activity limits for DAW (indoors/outdoors, elevated/ground release point) and for surface contamination in different fire zones. The scenarios were appropriate and were not tied to a particular operational state or configuration of the reactor building and systems.

Decommissioning activities will require ingress and egress from the Reactor Building and some systems may need to be deactivated (e.g., building purge) as decommissioning progresses. In addition, contaminated materials will be moved. Release points can be important as the atmospheric dispersion factors for elevated release are smaller than for ground release (7.67 x 10-4 compared to 1.52 x 10-3 s/m3). The Licensee addressed the NRC staffs concern with respect to release scenarios in the response to the NRC staffs RAI.

The NRC staff used GoldSim to verify the Licensees calculations, and some aspects of the Licensees calculations were sufficiently described and could be verified. However, in NRCs initial review some aspects could not be verified and therefore staff provided RAI 3. The amount of radioactivity released in response to a hypothetical fire was not clear. Some radionuclides (elements) were apparently assumed to be 100 percent available where others were 1 percent available and different ARFs were used for different elements. In addition, the total activity (Ci) of each radionuclide was provided by the Licensee along with the supporting references, but the source reference of that information could not be identified and obtained. In response to RAI 3, the Licensee performed the calculations described in the previous section. The NRC staff reviewed those calculations and determined that they have sufficient detail. However, a few minor discrepancies were identified:

• The table provided as an example on page 3 of the January 27, 2023, submittal did not match Table 17 from the supporting reference.

• The results provided for scenarios S1 and S2 in Table 12 of 164090-EN-CAL-004 Rev 0 are reversed.

The NRC staff calculated CEDE and CDE bone doses for the Licensees inventory mix that were within 1 percent of the Licensees values. The minor discrepancies noted above do not invalidate the conclusions made in this safety evaluation report.

The radiological inventory is important because it is the starting point or initial condition for estimating potential offsite doses resulting from accidents. In RAI 4, the NRC staff asked for additional information to support the radiological inventory present in the Reactor Building basement that formed the basis for the limiting scenario in previous analyses. The amount of some radionuclides (e.g., Pu-241) included in the fire analysis inventory were lower and not consistent with the results of sampling of the basement sediment (GEND-INF-011, Volume 3, 1983). In addition, the enrichment of the fuel at the time of the 1979 accident was not precisely

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known. While a higher enrichment is conservative with respect to criticality analysis, a lower enrichment is conservative with respect to analysis of a potential fire. In response to RAI 4, the Licensee indicated that the estimate of record for the fuel as UO2 remaining in the Reactor Building basement is 1.3 kg. The 1.3 kg of fuel would contain approximately 0.0264 g of Pu-241.

At a specific activity of 102.3 Ci/g, that corresponds to 2.7 Ci of Pu-241, which is less than the 10 Ci used in the Licensees LAR analysis. Data from A Data Report, Reactor Building Radiological Characterization, Vol 1. (TPO/TMI-125 Rev 2 (1989), non-public) suggests that the amount of Plutonium and Americium in solids removed from the Reactor Building b asement was even lower, suggesting the values used were conservative.

There is uncertainty in the radiological inventory, where it is located, what form it is in, and how available the material would be for release in an accident. Each measurement method, while complementary, has different types of challenges. The sediment samples were low volume and difficult to obtain, therefore they may not be representative. The smear type samples represent removable surface contamination for locations that are access ible. Difficult to reach areas may have higher amounts of removable contamination because they have not been accessed by remediation activities completed to date (e.g., defueling). The distribution of different radionuclides is not uniform throughout the reactor building and systems. The 1979 accident involved complex physical and chemical processes. The NRC staff considered these uncertainties when performing the review of the Licensees analysis.

The previous analyses of a Reactor Building fire assumed a hypothetical fire potentially involving all the residual radioactivity present in an area of the building (e.g., Reactor Building basement). The Reactor Building is very large with a relatively limited amount of combustible material present. The hypothetical large fires considered were unnecessarily conservative and not consistent with the proposed decommissioning approach. Conservative and hypothetical calculations can be useful in the regulatory process if they do not severely impact the desired actions, but it is appropriate to revisit and reassess those calculations if needed. In the second revised analysis to assess (limit) the impacts of a potential area fire, the Licensee proposed the use of management controls to ensure that a fire in the Reactor Building during decommissioning could be limited to an area of no greater than 930 m 2 (10,000 ft2).

The Licensee completed drone surveys and collected smear samples at a variety of locations to estimate the amount of removable surface contamination present throughout the Reactor Building. Using an assumed smear surface area of 100 cm2 (16 in2) and a collection efficiency (fractional) of 0.10, the Licensee measured Cs-137 activity calculated to correspond to a removable contamination of Cs-137 on an average of 2.35 x 10-6 Ci/ft2. Based on this surface activity, and the assumed MUP mixture (proportion of different radionuclides), the Licensee calculated that, on average, each 10,000 ft2 area evaluated would only contain 4 percent of the radioactivity necessary to exceed 2 times the ODCM limit to the limiting organ (bone). This percentage would be approximately 0.16 percent if a 1 mrem CEDE limit was used instead.

These area calculations rely on the appropriateness of the assigned radionuclide mixture. The Licensee stated that a Defueling mixture would be more conservative, but it was only representative for 3 of 47 areas examined. The radionuclide mixture assigned by the Licensee was the MUP mixture. For a Defueling mixture, over 90 percent of the inhalation dose is from Plutonium (Pu) and Americium (Am). For the MUP mixture, about 45 percent of the inhalation dose is from Pu and Am. The MUP mixture is depleted in Americium and Plutonium relative to Cesium compared to the Defueling mixture. Although the NRC Staff finds that the Licensees use of the MUP mixture was appropriate, those areas better represented by a Defueling mixture should have the specific management controls identified in the Licensees submittal to limit

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potential combustion sources in those areas to account for uncertainties.

The Licensees proposed approach assumed that radioactivity potentially released in a fire would be present in the removable contamination. Much of the radioactivity remaining in the Reactor Building or other buildings is contained within systems such as piping, drains, sumps, embedded in concrete, or present in sediment in the Reactor Building basement. The NRC staff finds that it was appropriate for the Licensee to evaluate removable contamination. However, some radioactivity is likely attached or absorbed to the combustible materials remaining in the Reactor Building. In addition, due to the radiation fields involved, the smear samples are biased to be from areas that have lower amounts of removable contamination as the samples are collected remotely. To evaluate the health and safety impacts of these uncertainties, the NRC staff calculated the inventory necessary to be present in the reactor building to produce an offsite dose that would exceed the EPA PAG of 1 rem in the event of a large fire. The NRC staff used an ARF of 5 x 10-4, otherwise the other parameters described previously by the licensee were used. The release was elevated with no HEPA filtration. The NRC staffs results are shown in Table 4 below. The only radionuclides with sufficient inventory (according to the 1990 inventory estimate) to by itself exceed the EPA PAG were Pu-239 and Am-241. This hypothetical bounding analysis is assuming that the total inventory is potentially available for release in a fire, which is not a credible event. There is not sufficient sources of fuel and combustible materials for a fire involving the whole Reactor Building, and most of the Reactor Building and internal components are made of non-combustible materials. For example, for a layer of sediment that is many centimeters thick and composed of primarily oxide minerals, only the radioactivity in the topmost layer is potentially available for release in a fire. In other words, the use of an ARF 5 x 10-4 (which the NRC staff used) applied to a layer of soil is extremely conservative.

A key parameter associated with estimation of offsite doses from an onsite release to the atmosphere is the atmospheric dispersion factor (C/Q). TMI-2 Solutions developed a value of 7.67 x 10-4 s/m3 for an elevated release and 1.52 x 10-3 s/m3 for a ground release. The values were developed based on an assumed atmospheric stability class of F and a constant wind speed of 1 m/s. In previous licensing documents the C/Q value was inaccurately described as worst case (ML21057A046). This parameter is site specific and depends on the atmospheric conditions at a particular site. The values are not worst case, but do represent conditions that would only be worse (from an offsite dose calculation perspective) roughly 5 percent of the time.

The NRC staff concludes that the values selected by the Licensee in the LAR were appropriate to use in the accident analysis.

In RAI 5, the NRC staff asked TMI-2 Solutions for additional information associated with the management and control of the buildup of radiolytic gases that could pose an explosion hazard.

Interaction of radiation with water or other materials can result in the production of radiolytic gases, primarily hydrogen (H2). In sufficient concentrations and with oxygen present, hydrogen is flammable. Through operation of the Submerged Demineralizer System and packaging of the generated waste for disposal, it was observed that debris material could generate H2 inshort-term storage that could reach Lower Flammability Limits (LFL). Licensing of the dry cask storage system in Idaho for the debris material applied multiple controls and systems in order to prevent buildup of H2 gas to the LFL (ML18296A527). Canisters were vacuumed dried prior to storage and the systems included a HEPA filter to vent hydrogen. Monitoring of hydrogen levels is performed (ML19259A017) and observed hydrogen levels have been around 0.04 percent where the LFL with oxygen present is 5 percent - the venting has been very effective but hydrogen generation is continual. Though a large fraction of the radioactivity has been removed from the TMl-2 systems and liquids have been removed, high radiation fields and potential

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liquids remain. The deactivated reactor systems have dead end and closed portions (e.g., high points in unused piping) where H2 gas could collect. Potential moisture is present in systems and components. Decades have passed since the accident where H2 could be generated. Then licensee GPUN indicated that, in preparation for entry into PDMS, the plant systems were vented, drained, and the remaining water volumes were processed for disposal. However, GPUN recognized that there could be small pockets of hydrogen gas remaining in some portions of plant systems and components. TMI-2 Solutions proposed the establishment of work planning instructions in work packages which would evaluate specific hydrogen concerns relevant to the given scope of work and include appropriate hydrogen mitigation measures appropriate for that work. The NRC staff finds that the Licensees proposal is appropriate to mitigate the hazard of radiolytically generated hydrogen.

Table 4 Radionuclide Inventory to Exceed EPA PAGs for an Extreme Fire Radionuclide Inventory 1990 (Ci) Limiting Inventory (Ci)

Sr-90 2,400 5,700 Cs-137 43,000 230,000 Am-241 22 17 Pu-239 90 24 Pu-241 950 1,500

In RAI 6, the NRC staff asked the Licensee to provide information on dust explosions and exothermic reactions. The licensees response (ML20127L978) provided sufficient basis that dust explosions or exothermic reactions are unlikely to pose a hazard during decommissioning.

Use of proper housekeeping can help ensure a dust explosion will not occur. It is unlikely that reactive (unoxidized) particles remain in the systems and components. Because the processes during the reactor accident were complex, it is recommended that the licensee communicate the potential hazard of disturbing material that could remain reactive to staff who will perform the active decommissioning work so that unforeseen events or conditions can be properly assessed and mitigated (if they were to occur).

The licensee provided additional information associated with a fire of the cork seams (RAI7).

The licensees response was appropriate and provided adequate basis that a fire involving the contaminated cork seams was unlikely.

The licensee appropriately considered other events that could potentially exceed the threshold of an NOUE that had previously been assessed in a variety of documents. The events that could potentially 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. NRC staff agrees that these events were appropriate and sufficiently comprehensive. The licensee stated that they have no plans to use oxyacetylene in the vicinity of high source term components, instead they will use mechanical means for size reduction. Use of oxyacetylene would be controlled by the TMI-2 Engineering Program. With respect to a drop of a spent resin liner, a decision was made to locate the proposed water processing system in the HEPA filtered FHB and thus eliminate the scenario for an outside unfiltered release. Limits on the allowable radioactivity in the waste processing tanks were previously established and reviewed by the NRC.

The Licensee made a variety of commitments associated with management controls and practices that would be used to mitigate uncertainties and ensure that public health and safety would be protected. Specifically, the Licensee committed to establish work instructions in the

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form of work packages which will evaluate specific hydrogen concerns relevant to a given scope of work and include appropriate hydrogen mitigation measures appropriate for that work (ML22276A024, ML23049A004); committed to place limits on the activity content of the combustibles available for a fire to limit the severity and to prevent the release of radioactive above established criteria (ML22276A024); and committed to use the established Source Term Limitations and Administrative Controls for the TMl Station Decommissioning Emergency Plan Action Levels which establishes limits on the accumulation of uncontained radioactive materials to ensure the criteria for the NOUE will not be exceeded in the event of a postulated fire (ML23033A103). The NRC determined that the following license conditions in the form of TS (Section 6.0, Administrative Controls), are necessary because the administrative safety commitments should be part of the licensing basis. The Licensee agreed to the license conditions and their location in TS Section 6.0 (ML23090A214) below:

TS 6.15.1:

Establish work planning instructions in work packages to include appropriate hydrogen evaluation and mitigation measures for activities involving debris material removal from Systems, Structures and Components.

TS 6.15.2:

Establish Source Term Limitations and Administrative Controls for activities involving the removal and handling of combustible radioactive material to ensure the criteria for Notification of an Unusual Event emergency classification of two times the Offsite Dose Calculation Manual dose limits will not be exceeded in the event of a postulated fire.

Because the Licensee will be using the established Source Term Limitations and Administrative Controls for the TMl-2 Decommissioning Emergency Plan Action Levels, which establish limits on the accumulation of uncontained radioactive materials, the criteria for the NOUE will not be exceeded in the event of a postulated fire for activities involving debris material removal and handling.

The NRC staff examines the information provided by a licensee in a risk-informed context when making a regulatory decision. The Licensee purposely introduced conservatisms in a variety of parameter assignments and assumptions in the accident analysis. The approach to accident analysis in unto itself can also introduce conservatism. For example, the figure below shows the Reactor Building and a 600 m (2000 ft) radius from that point. The analysis assumes that the receptor would be located at the EAB for the duration of the fire accident. While it is possible that a receptor would be located at the TMI-2 site boundary for the duration of an accident, based on the current population distribution of the immediate area, that is unlikely. The limiting receptor is likely to be more distant from the TMI-2 Reactor Building which would result in greater atmospheric dispersion. As previously discussed, the use of a limit that is twice the limiting organ dose for the ODCM is very restrictive from a radiological protection viewpoint. A release from a fire is not a normal effluent release, the release is likely to be of short duration and dissipate rapidly.

The NRCs approval of the derived operational values does not mean that individual parameter values are appropriate for use in a different licensing context. For example, not every parameter assigned in the Licensees analysis was conservative. The NRC considers all available information including conservatism and non-conservatism. The NRC staff finds that the Licensees use of conservative values in many key inputs provides reasonable assurance that

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the derived operational limits will ensure that offsite public dose requirements will be met during decommissioning. For these types of analyses, use of a probabilistic calculation would allow the direct consideration of uncertainty and variability of key parameters and a relaxation of the conservatism introduced in many of the parameters used in the analysis to account for uncertainty.

Figure 1 A 2,000 ft Radius from the TMI-2 Reactor Building

5.1.1 Conclusion

The NRC staff concludes that the Licensees proposed exit from PDMS and entrance into DECON will not result in any credible accident scenarios that will exceed the EPA PAG of 1 rem and will not 1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The Licensees proposed activity limits for DAW and HEPA filters will ensure that credible accidents that result in the release of radioactivity to a member of the public at the EAB exceeding the NOUE criteria is unlikely to occur. The NRC staff made the licensees commitment part of the licensing basis by adding TS, as discussed in this SER and agreed upon by the Licensee (ML23090A214).

5.1.2 References

1. NUREG/CR-26-01, Vol.1 of 2, Technology, Safety and Costs of Decommissioning Reference Light Water Reactors Following Postulated Accidents. (ML14023A049)

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2. TMI-2 ISFSI Renewed Technical Specifications [Letter to J. Zimmerman re: Issuance of Renewed Materials License No. SNM-2508 for the Three Mile Island Unit 2 Independent Spent Fuel Storage Installation, 9/16/2019, ML19259A017.

3. Three Mile Island, Unit 2, License Amendment Request Decommissioning Technical Specifications, 2/19/2021, ML21057A046.

4. Safety evaluation supporting amend 48 to license DPR-73, 12/28/1993, ML20059D177.

5. License Amendment Request - Three Mile Island, Unit 2, Decommissioning Technical Specifications, Supplemental Information, 1/7/2022, ML22013A177.

6. TMI-2 PDMS SAR Update 14 & QA Plan (Rev 18 & 19) Biennial Submittal 08-10-21, 8/10/2021, ML21236A288.

7. Programmatic Environmental Impact Statement related to Decontamination and Disposal of Radioactive Wastes Resulting from March 28, 1979 Accident Three Mile Island Nuclear Station, Unit 2, Final Supplement, 8/31/1989, ML20247F778.

8. Requests for Addition Information for Proposed Decommissioning Tech Specs License Amendment Request, 8/5/2022, ML22210A080.

9. 164090-EN-CAL-004 (Proprietary), 2/11/2021.

10. NEI 99-01 Revision 6, Development of Emergency Action Levels for Non-Passive Reactors,11/21/2012, ML12326A805.

11. License Amendment Request - Three Mile Island, Unit 2, Decommissioning Technical Specifications, Response to Request for Information Regarding Radiation Protection Program Ventilation Controls,11/7/2022, ML22313A050.

12. Forwards Rev 2 to "SER for Early Defueling of TMI-2 Reactor Vessel, 5/20/1985.

13. REDACTED: TMI, Unit 2, Independent Spent Fuel Storage Installation -

Application for 10 CFR 72 Specific License Renewal, 9/26/2018, ML18296A527.

14. GEND-031, Submerged Demineralizer System Processing of TMl-2 Accident Waste Water, February 1983.

15. U.S. Nuclear Regulatory Commission (NRC). 1981. Final Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Waste Resulting from March 28, 1979, Accident at Three Mile Island Nuclear Station, Unit 2. NUREG-0683, U.S. Nuclear Regulatory Commission, Washington, D.C., ML19343C359.

16. Letter C312-91-2023 from Long R. L. (GPU Nuclear), Post-Defueling Monitored Storage Safety Analysis Report Amendment 11, April 19, 1991 (9106190266.pdf)

17. Letter Masnik, M.T. (NRC) Proposed Possession Only License, Proposed Technical Specifications and Supporting Safety Evaluation for Post Defueling Monitored Storage at Three Mile Island Unit 2 (92003040090)

18. DOE, Airborne Release Fractions/Rates and Respirable Fractions for non-Reactor Nuclear Facilities, DOE-HDBK-3010-94, US Department of Energy, Washington DC, 1994, ML13078A031.

19. Cox, T.E., J.T. Horan, G. Worku, Reactor Building Basement Radionuclide and Source Distribution Studies, GEND-INF-011, Volume 3, EG&G Idaho, Idaho Falls ID, 1983.

20. Smith, R.I., G.J. Konzek, and W.E. Kennedy, Jr., Technology, Safety,

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and Costs of Decommissioning a Reference Pressurized-Water Reactor Power Station, (Report prepared by Pacific Northwest Laboratory, Richland, Washington), NUREG/CR 0130, Vol. 1 and 2, U.S. Nuclear Regulatory Commission, June 1978.

21. TPO/TMI-125, Data Report, Reactor Building Radiological Characterization, Volume 1, December 1989.

5.2 Nuclear Criticality Safety - Updated Safe Fuel Mass Limit

The NRC staff reviewed the Licensees updated SFML relative to the associated activities that are to be conducted via the 10 CFR 50.59 process involving the remaining debris material to determine whether subcriticality is assured under normal and all credible abnormal conditions.

The TMI-2 reactor has been undergoing under various stages of defueling and cleanup activities since 1979 through 1993, with the primary defueling activities occurring in the late 1980s.

Defueling activities originally focused on the removal of fissionable material from the reactor vessel (RV), removing approximately 99 percent of the original core inventory. This was accomplished using a safe fuel mass limit (SFML), which was established based on conservative assumptions.

An estimated 1097 kilograms (kg) of residual UO2, ~1.2 percent of original TMI-2 inventory, is still present within the RV and in various locations outside of the RV. To support the defueling of the remaining fissionable material, TMI-2 Solutions established an updated SFML based on more realistic conditions and sampling data. The new SFML intends to bound all remaining activities including removal of debris material from the RV, movement to the reactor cavity or other area intended for segmentation, and movement to loading of the transportable storage container (TSC).

The remaining debris material is in the form of finely divided, small particle-size sediment material; resolidified material either tightly adherent to components or in areas inaccessible to defueling at the time of the initial clean up after the 1979 accident up until the completion of the defueling8; and adherent films on surfaces contained within piping, tanks, and other components. The debris material is present both inside and outside of the RV, with most of the mass residing in the lower head of the RV. Licensee evaluations of the residual fuel suggests that no discrete (i.e., neutronically de-coupled) location has in excess of 127 kg UO2, and the total estimate of mass present outside of the RV is 170 kg UO2.

5.2.1 Description of Normal and Credible Abnormal Conditions

Under anticipated normal conditions, segmentation and loading of TSCs will occur in the reactor cavity where material is transported to after it is removed. Most of the component segmentation will occur under water in a flooded reactor cavity; however, some segmentation may not occur under water, but rather on a concrete platform. Material and component removal from each area will be treated as separate operations to ensure that debris material from two separate areas will not be removed at the same time; however, a partially filled TSC from one area may remain for loading from the next area if there is remaining void space.

Credible abnormal conditions associated with decommissioning operations may include the mishandling of components with fissionable material to increase moderation, reflection

8 See TMI Nuclear Station Unit 2 Defueling Completion Report (Ref.3)

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conditions, or interaction. This could potentially occur through the removal and accumulation of fissionable-bearing components from multiple areas into a single location or through the accumulation of fissionable-bearing segmented pieces in the Waste Zone (e.g., segmented pieces that do not fit into the waste basket liner that have been set aside).

5.2.2 Nuclear Criticality Safety Analysis

The Licensee performed a set of nuclear criticality safety (NCS) calculations using Monte Carlo N-Particle (MCNP) 6.2 with the (Evaluated Nuclear Data File (ENDF)/B-VIII.0 cross section library to bound the normal conditions of its proposed activities. Credible abnormal conditions are discussed in Section 5.2.3 of this report. The details of the Licensees validation of MCNP 6.2 are discussed in Section 5.2.4 of this report. The bounding configurations included the following:

• Optimally-moderated, fully-reflected heterogeneous sphere with impurities (fuel and impurities in a hexagonal lattice contained in a spherical geometry);

• Optimally-moderated, fully-reflected heterogeneous sphere with poisoned moderator (fuel in a hexagonal lattice contained in a spherical geometry); and

• Optimally-moderated, fully-reflected heterogeneous sphere with impurities and poisoned moderator (fuel and impurities in a hexagonal lattice contained in a spherical geometry).

For each configuration evaluated, the Licensee applied the following assumptions:

• Fuel pellet volume equivalent to a standard size fuel pellet, but in a spherical geometry and in a hexagonal lattice;

• Uniform, homogeneous distribution of impurities (when considered) within the fuel sphere;

• Full, tight-fitting (essentially infinite) water reflection;

• Fuel composition of TMI-2 average fuel (i.e., a homogeneous mixture of the three TMI-2 fuel batches);

• No credit for existing structural or solid poison materials (e.g., borosilicate glass); and

• Limited credit for the presence of fission product poisons.

The NRC staff reviewed the Licensees analysis and associated assumptions for the key parameters of UO2 density, geometry, heterogeneity, fissionable material composition, impurity composition, moderation, and reflection. The NRC staffs evaluation for each parameter is discussed below.

• Density (UO2). The Licensee considered the full theoretical density of UO2 (10.97 g/cm3),

a reduced density based on burn -up history (10.55 g/cm3), and the as-built density of 10.14 g/cm3. The Licensee determined that changes to UO2 density resulted in a minimal effect on reactivity, and the Licensee ultimately used the full theoretical density in the establishment of its updated SFML. Although the UO2 density may vary throughout the remaining debris material, the use of full theoretical density is bounding. Therefore, the NRC staff considers the use of full theoretical UO2 density to be conservative and bounding for the normal and credible abnormal conditions of the proposed activities.

• Geometry. The Licensee assumed a fuel and impurities (fuel/impurities) hexagonal lattice contained within a spherical geometry. Fuel pellets were also assumed to be spherical in a hexagonal lattice. The exact geometry of the fuel is not well-known and

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may vary spatially throughout; therefore, the NRC staff considers the use of a bounding geometry appropriate. In general, spherical geometries are bounding due to the effects of surface area-to-volume-ratio on neutron reflection. The use of a hexagonal lattice affords the ability to closely pack fuel elements, which maximizes the range in which interspersed moderation can be assessed ( Moderation is discussed below). Therefore, the NRC staff determined that the Licensees use of a hexagonal lattice and spherical geometry for both fuel pellets and the fuel/impurities mixture was conservative and bounding for the normal and credible abnormal conditions of the proposed activities.

• Heterogeneity. The Licensee assumed a heterogeneous sphere to model the fuel configuration. The exact configuration of the fuel is not well-known and may vary spatially throughout; therefore, the NRC staff considers the use of a bounding assumption with respect to heterogeneity is appropriate. In many cases, homogeneous mixtures represent the most reactive configuration; however, this is not necessarily true in some cases (e.g., certain low-enriched uranium systems). The NRC staff reviewed scoping calculations in NSTS-ES-01, Safe Fuel Mass Limit Criticality Safety Analysis for TMI-2 Decommissioning (Ref. 7) and performed independent scoping calculations to assess whether a heterogeneous system is bounding (primarily to a homogeneous system) in this case. The NRC staff determined that a heterogeneous model wa s more reactive than a homogeneous model and, therefore, bounds both heterogeneous and homogenous systems. Therefore, the NRC staff determined that the Licensees use of a heterogeneous model was conservative and bounding for the normal and credible abnormal conditions of the proposed changes and the activities conducted under 10 CFR 50.59 involving debris material.

• Fissionable Material Composition. The Licensee represented the fissionable material as a homogeneous medium. The enrichment of the fuel was assumed to be the average of the three known fuel batches for TMI-2. The unburned fuel enrichments for the three batches were 1.98 weight percent (wt. %) (56 assemblies), 2.64 wt. % (61 assemblies),

and 2.96 wt. % (60 assemblies), which equates to an average enrichment of 2.54 wt. %

235U/U. The TMI-2 fuel experienced the equivalent of approximately 94 effective full-power days of burn-up at the time of the accident. The actual exposure history for each fuel batch, using existing plant data, was applied to estimate burn-up effects. The exposures and core operating history were applied to a SCALE/ORIGEN-S model to calculate isotopic inventory at the time of the accident. The incorporation of the burn-up effects for each fuel batch was used to produce a net enrichment of 2.24 wt. % 235U/U for a homogeneous mixture of the three batches. The starting fuel composition was decayed using SCALE/ORIGEN to the year 2022.

The NRC staff reviewed the Licensees calculations, the sampling data detailed in the TMI Nuclear Station Unit 2 Defueling Completion Report (Ref.3), and the information contained in TMI2-EN-RPT-0001, Determination of the Safe Fuel Mass Limit for Decommissioning TMI-2 (Ref. 1), to evaluate whether the Licensees assumptions regarding fissionable material composition, enrichment, and homogeneity were acceptable. Additionally, the NRC staff performed an independent confirmatory analysis using the S CALE/ORIGEN package. With respect to composition and burn-up, the NRC staff observed via its independent analysis that the decayed fissionable material composition was more reactive than that of the original burn-up composition. The staff determined that this was primarily due to the decrease in europium (Eu)-

154 and 155Eu abundance. The NRC staff determined that the Licensees assumptions regarding fissionable material burn-up were acceptable and more conservative than the

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assumptions used for the Licensees original SFML. With respect to fissionable material enrichment, the NRC staff confirmed via its independent analysis that the Licensees calculated net enrichment of 2.24 wt. percent 235U/U for a homogeneous mixture of the three fuel batches was acceptable. The NRC staff noted that the 34 samples obtained from the bottom head of the RV had an average enrichment of 2.23 wt. percent 235U/U, which is in good agreement with the Licensees calculated net enrichment of 2.24 wt. percent 235U/U. The NRC staff also noted that defueling data suggests that approximately 65 percent of the batch 3 fuel, which had an initial enrichment of 2.96 wt. percent 235U/U, has already been removed from the RV. Therefore, the remaining fuel consists primarily of batches 1 and 2 with enrichments of 1.98 wt. percent and 2.64 wt. percent 235U/U, respectively. This provides additional assurance that the Licensees calculated net enrichment of 2.24 wt. percent 235U/U is bounding and acceptable.

With respect to homogeneity, a substantial amount of core debris material was removed during defueling in the late 1980s. Prior to the defueling, a sampling of core debris material from the center and radius was performed, removing material from three depths: 1) surface of the rubble bed; 2) three inches deep into the rubble bed; and 3) twenty-two inches deep into the rubble bed. The samples showed a high degree of homogeneity, and the degree of mixing and relocation of core constituents suggested that molten core materials were mixed vigorously. The samples indicated that most of the debris material in the RV was mixed to create a degree of homogeneity as evidenced by the narrow range of uranium and zirconium (Zr) composition observed in obtained samples (62.3 percent - 69.5 wt. percent U with an average of 65.1 wt.

percent, and 11.7 wt. percent - 15 wt. percent Zr with an average of 12.6 wt. percent) and the distribution of impurities (the distribution of impurities is discussed below in the Impurity Composition section). The ratio of uranium-to-zirconium (U/Zr) for the obtained samples was observed to be higher (~5.9) than that of as-built (~3.6), providing further evidence of significant mixing of all core constituents. The NRC staff considers the samples to be reasonably representative of the remaining debris material in TMI-2 based on the numerous locations sampled and potential fuel relocation pathways. Therefore, the staff considers it unlikely that areas containing exceedingly high uranium concentrations exist. The NRC staff determined that the Licensees assumption that the fissionable material is homogenous is acceptable.

Impurity Composition. Fissionable material, control rods, structural materials, and other miscellaneous materials were melted and mixed during the 1979 accident. The remaining material contains both fissionable material and non-fissionable materials (i.e., impurities) from structural materials, control rods, burnable poison rods, coolant poison, and cladding. Core debris material in the lower head region of the RV is most representative of what remains in the RV at the present time. For this reason, the Licensee based its assumptions for impurity composition on samples obtained from the lower head region of the RV as detailed in TMI Nuclear Station Unit 2 Defueling Completion Report (Ref. 3). Based on these samples, the fuel composition (UO2 + fission products) was assumed to be 83.79 wt. percent, while the remainder (16.21 wt. percent) was assumed to be impurities.

Table 3.2.2.2 Impurity Composition Component Weight Percent (wt. %)

Fuel Composition 83.79 Zirconium (Zr) 12.70 Iron (Fe) 2.44 Boron (B) 0.11 Cadmium (Cd) 0.00 Chromium (Cr) 0.75

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Molybdenum (Mo) 0.15 Manganese (Mn) 0.06

TMI-2 Solutions performed a series of calculations varying the impurity composition from 0 percent to 16.21 percent with a fixed iron content of 90 percent (of the relative impurity mixture) and varied boron content. The Licensee determined that it was necessary to apply partial credit to the impurity composition, taking credit for 10 wt. % of the impurities, 50 percent of the sampled boron content of those impurities (i.e., 5 percent of the overall boron content), and 90 percent of the iron content (i.e., 45 percent of the overall iron content).

The NRC staff reviewed the Licensees calculations, the sampling data detailed in the TMI Nuclear Station Unit 2 Defueling Completion Report (Ref.3.) and the information contained in TMI2-EN-RPT-0001, Determination of the Safe Fuel Mass Limit for Decommissioning TMI -2 (Ref. 1) to evaluate whether the Licensees assumptions regarding impurity composition were acceptable. The NRC staff noted that samples obtained from the RV and B steam generator tube sheet showed the presence of impurities, and the samples indicated that m ost of the debris material in the RV was mixed to create a degree of homogeneity as evidenced by 1) the narrow range of uranium and zirconium composition observed in obtained samples (62.3 percent - 69.5 wt. percent U with an average of 65.1 wt. percent, and 11.7 wt. percent - 15 wt. percent Zr with an average of 12.6 wt. percent); 2) the presence of gadolinium (Gd) in samples obtained from the core debris material bed; 3) presence of gadolinium at a location at least 30 centimeters from its original position in the TMI-2 core; and 4) the consistent, evenly-distributed and relatively stable presence of boron in all of the lower head samples. The NRC staff considers the samples to be reasonably representative of the remaining debris material in TMI -2 based on the numerous locations sampled and potential fuel relocation pathways. Therefore, the NRC staff determined that the Licensees assumption that impurities are distributed homogeneously is acceptable. The NRC staff also noted that homogenization of the impu rity content of the RV lower head (83.79 wt. percent of the TMI-2 core mass) represents a significant conservatism over what would be derived assuming a homogeneous mixture of the initial core composition, where the UO2 component is 65.8 wt. percent of the TMI-2 core mass. This provides additional conservatism with respect to the Licensees impurity composition assumptions.

With respect to applying partial credit to the boron and iron content of the impurities, the NRC staff performed a sensitivity study to assess the impacts of varied boron concentration. This study was performed due to boron having the highest (with respect to impurities) neutron absorption cross section for the energy region of interest and greatest impact on keff. The NRC staff observed that over the range of 0 - 100 percent boron impurity, the average worth was -

0.48 percent Dk/k for each 1 percent addition and that as boron impurity increases, negative reactivity worth decreases. At 10 wt. percent impurity content, 20 percent of the boron impurity is worth -11.4 percent Dk/k, which equates to -0.57 percent Dk/k for each 1 percent addition.

The NRC staff determined that for a 1200 kg U fuel loading (i.e., the proposed SFML) taking credit for 10 percent of the measured impurity content, 47 percent of the boron content must be credited to stay below the Licensees upper subcritical limit (USL) of 0.95. This corresponds to an overall boron content of 0.0319 wt. percent of the remaining debris material.

Samples indicate that boron is deposited at a relatively constant concentration evenly throughout the lower head and is consistent with the amount of boron present in the reactor coolant. For iron, samples indicate that substantial amounts of structural material have been retained in the debris material, with iron having the highest concentration and being relatively constant throughout (1.8 - 3.7 wt. percent Fe with an average of 2.44 wt. percent). As

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previously stated, the NRC staff considers the samples to be reasonably representative of the remaining debris material in TMI-2 based on the numerous locations sampled and potential fuel relocation pathways. Therefore, the NRC staff considers it unlikely that large areas exist within the debris material in which little to no boron or iron is present. Even if this were to occur, the Licensees SFML relies on only 10 wt. percent of the impurity content with credit for boron and iron content both further reduced by 50 percent and 90 percent, respectively. This corresponds to a credit of only 5 percent of the measured boron content and 45 percent of the measured iron content. Furthermore, as discussed in the Fissionable Material Composition section above, the NRC staff considers it unlikely based on sampling data that areas exist in which exceedingly high uranium concentrations exist. Therefore, the NRC staff considers it unlikely that an area in which little to no boron or iron exists, and the NRC staff considers it further unlikely that such an area would also be of high uranium concentration. For these reasons, the staff determined that the Licensees assumptions and application of partial credit for boron and iron content (10 wt.

percent of the impurities, 50 percent of the sampled boron content of those impurities, and 90 percent of the iron content) is acceptable, conservative, and bounding to any uncertainty associated with the impurity content. The NRC staff noted that borosilicate glass was added to TMI-2 in the 1990s, which was not considered as an impurity or credited, but would serve to neutronically isolate any debris material that is located or could relocate to the bottom of the RV (Refs. 10 and 11). This provides additional conservatism as the borosilicate glass further reduces the reactivity of the system. The NRC staff also noted that although gadolinia burnable poison rods were used in TMI-2 and gadolinium, therefore, may be present in certain locations (e.g., core debris material bed), the presence of gadolinium (a neutron absorber) was not credited.

Moderation. The Licensee performed a series of calculations to optimize the moderator-to-fuel ratio by varying the fuel pellet diameter and lattice pitch (i.e., hydrogen-to-uranium ratio). The optimized moderator-to-fuel ratios determined by these calculations were used for each case in the establishment of the SFML.

The NRC staff reviewed the Licensees series of calculations to optimize the moderator-to-fuel ratio and performed an independent confirmatory analysis. Many of the proposed activities will take place in the presence of significant amounts of moderating material (i.e., underwater), and the exact configuration of the fuel and moderating material is either not well-known or may change during the proposed activities. Therefore, the NRC staff determined that the Licensees use of optimized moderator-to-fuel ratios was appropriate. The NRC staff confirmed via its independent confirmatory analysis that the values the Licensee identified as optimum represented worst-case moderation conditions and were, therefore, acceptable and bounding for the normal and credible abnormal conditions of the proposed activities.

Reflection. TMI-2 Solutions considered full, tight-fitting reflection for all cases. The Licensee also considered concrete as a reflector.

Many of the proposed activities will take place in the presence of significant amounts of reflecting material (i.e., underwater), and the exact configuration of the fuel and reflecting material is either not well-known or may change during the proposed activities. Additionally, certain activities may take place on a concrete slab. Therefore, the NRC staff determined that the Licensees use of full, tight-fitting reflection and consideration of concrete was acceptable and bounding for the normal and credible abnormal conditions of the proposed changes and associated activities that are to be conducted via the 10 CFR 50.59 process involving the remaining debris material.

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Results. The results of the licensees analysis established a new SFML of 1200 kg U (1361 kg UO2), which takes credit for fuel exposure histories and 10 wt. percent of the impurities identified in the TMI Nuclear Station Unit 2 Defueling Completion Report (Ref. 3), with boron and iron further reduced by 50 percent and 10 percent, respectively. Based on the derived SFML, which is greater than the estimated remaining 1097 kg UO2, the licensee determined that all associated activities that are to be conducted via the 10 CFR 50.59 process involving the remaining debris material can be safely conducted, and that no credible criticality hazard exists.

5.2.3 Credible Abnormal Conditions

Credible abnormal conditions associated with decommissioning operations may include the mishandling of components with fissionable material to increase moderation, reflection conditions, or interaction. This could potentially occur through the removal and accumulation of fissionable-bearing components from multiple areas into a single location or through the accumulation of fissionable-bearing segmented pieces in the Waste Zone (e.g., segment ed pieces that do not fit into the WBL that have been set aside).

Given the size and complexity of the reactor components and the effort required to remove and relocate each component, the NRC staff considers it unlikely that components from multiple areas would be allowed to accumulate in a single location. Similarly, the NRC staff considers it unlikely that an accumulation of segmented pieces in the Waste Zone would occur. Still, if such accumulations were to occur (whether they be comprised of reactor c omponents or segmented pieces), the fissionable material is either adhered to, or non-uniformly distributed throughout, non-fissionable structural components. Therefore, the NRC staff determined that there is no credible upset that could lead to an accumulation of debris material mass in a geometrical configuration capable of criticality.

Because there is a large degree of uncertainty associated with the amount of fissionable mass in the RV (+/- 370 kg UO2), the total mass within the RV could be in excess of the licensees proposed SFML of 1200 kg U (1361 kg UO2). However, given the size of the RV and distribution of the debris material therein, it is not logistically possible for the entire fuel mass within the RV to be removed at one time. Furthermore, it is not possible for the entire fuel mass within the RV to be placed into a single TSC given the size restrictions of a TSC. Therefore, the NRC staff determined that there is no credible upset that could lead to an accumulation of fissionable mass in excess of the Licensees proposed SFML. As an additional defense in depth consideration, the fissionable material is either adhered to, or non-uniformly distributed throughout, non-fissionable structural components, making the formation of an unfavorable geometric configuration unlikely. Therefore, the NRC staff determined that there is no credible upset that could lead to an accumulation of debris material mass in a geometrical configuration capable of criticality.

5.2.4 Minimum Margin of Subcriticality, Upper Subcritical Limit, and Estimation of Bias

Because the Licensee relied on the use of computational methods (MCNP 6.2) to establish the SFML and demonstrate the no credible criticality hazard exists, the NRC staff reviewed the Licensees criticality code validation report, TMI-EN-RPT-0002, MCNP Version 6.2 Bias Determination for Low Enrichment Uranium Using the ENDF/B -VIII.0 Cross Section Library, to evaluate the Licensees proposed minimum margin of subcriticality (MMS) and associated USL.

The Licensee performed its validation using nine benchmark sets consisting of a total of 125 benchmark experiments to estimate a total bias (bias + bias uncertainty) of 0.0202, and the

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Licensee established a single USL of 0.95 and MMS of 0.0298 to bound the normal and credible abnormal conditions of all proposed decommissioning activities. The validation applied to heterogeneous spherical and annular low-enriched UO2 and water systems with an area of applicability (AOA) that includes various fissionable material forms, fissionable system geometries, moderating and reflecting materials, and neutron absorbers. The details of the AOA are discussed in Table 3.3.2.4-1.

Table 3.2.2.4 Validated AOA Property Area of Applicability Fissionable Materials Uranium (2 - 10 wt. percent) in the form of uranium dioxide (UO2), uranium tetrafluoride (UF4), and uranyl fluoride solution (UO2F2)

Geometry array of fuel rods, array of rectangular parallelepipeds, UO 2 in cubes in a cubic array, finely divided particles in cubes, cylinders, spheres, slabs Moderating Materials water, paraffin, polyethylene Reflecting Materials none, water, acrylic, steel, plexiglass, paraffin, polyethylene, concrete Absorbers, Poisons, and Structural Materials aluminum alloys, steel, borated steel, Boral, boroflex, silver-indium-cadmium, copper, copper with cadmium, cadmium, Zircalloy -4, rubber Average Energy of Neutrons Causing Fission Range: 2.43E 2.88E-01 MeV (AENCF) Average: 1.40E-01 MeV Bias + Bias Uncertainty 0.0202 MMS 0.0298 USL 0.95

The NRC staff determined that the validation of MCNP 6.2 was performed in accordance with American National Standards Institute/American Nuclear Society (ANSI/ANS) -8.24-2017, Validation of Neutron Transport Methods for NCS, NUREG/CR -4604, Statistical Methods for Nuclear Material Management, and NUREG/CR -6698, Guide for Validation of NCS Calculational Methodology. The Licensee performed the Lilliefors Test to determine whether the selected data were normally distributed and performed trending analyses for the parameters of average energy of neutrons causing fission (AENCF) and enrichment to determine whether any notable trends in the bias existed. Because the data were found to be not normally distributed with no apparent trends in the bias, the Licensee estimated the bias using the single-sided tolerance limit statistical method, which is consistent with NUREG/CR-6698.

The NRC staff determined that the Licensees methods for estimating bias and establishing the USL used technically sound statistical approaches consistent with NUREG/CR-6698 and provided a high degree of confidence that the selected experiments evaluated under these methods will yield an accurate estimate of the bias. The NRC staff determined that the selected experiments were drawn from multiple, independent series and a well-established source (the International Handbook of Evaluated Criticality Safety Benchmark Experiments, Ref. 9). The NRC staff determined that the selected experiments appropriately spanned the entire range of specified important parameters (fissionable material form and composition; geometry; enrichment, moderating, reflecting, and absorber materials form and composition; and AENCF) without large gaps requiring extrapolation or wide interpolation, and that the specified ranges

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included appropriate considerations for the potential values of important parameters under all normal and credible abnormal conditions. The NRC staff determined that a sufficient quantity of benchmark experiments (125 experiments) to draw a physically and statistically meaningful result was ensured through the use of statistical techniques. The NRC staff determined that the selected benchmark experiments were sufficiently similar to the normal and credible abnormal conditions presented by the Licensees associated activities that are to be conducted via the 10 CFR 50.59 process involving debris material and that they provide an acceptable level of assurance that the estimated bias is accurate and bounding to its various potential sources. The selected experiments were sufficiently similar for key parameters, and the key parameters used for benchmark experiment selection were those with the greatest effect on the bias, including fissionable material form and composition; enrichment; physical form and composition of moderating materials, reflecting materials, and neutron absorbers; and AENCF. The NRC staff determined that the MMS was generally large in comparison to the bias uncertainty. In aggregate, the NRC staff determined that the Licensees proposed MMS (0.0298) and associated USL (0.95) were acceptable for the purposes of conducting NCS analyses using MCNP 6.2 with the ENDF/B -VIII.0 cross section library. Licensees proposed MMS (0.0298) and associated USL (0.95) were acceptable for the purposes of conducting NCS analyses using MCNP 6.2 with the ENDF/B -VIII.0 cross section library.

5.2.5 Conclusions

The NRC staff determined that the Licensees associated activities that are to be conducted via the 10 CFR 50.59 process involving the debris material do not present any credible criticality hazards. Therefore, the NRC staff determined that the associated activities will not 1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. and that the associated activities that are to be conducted via the 10 CFR 50.59 process involving the debris material will be conducted such that subcriticality is assured under normal and all credible abnormal conditions. No changes to the POL or TS are required due to the proposed SFML.

5.2.6 - References

1. Revision to TMI2-EN-RPT-0001, Determination of the Safe Fuel Mass Limit for Decommissioning TMI-2.

2. M. Pritchard, TMI2-EN-RPT-0003, TMI-2 Safe Fuel Mass Limit Computational Input Consensus, Revision 0, Nuclear Safety and Technology Services, 2020.

3. GPU Nuclear Corporation, TMI Nuclear Station Unit 2 Defueling Completion Report, Revision 4, 1989.

4. ANSI/ANS-8.24-2017, Validation of Neutron Transport Methods for NCS Calculations, American Nuclear Society, La Grange Park, IL, 2017.

5. J. Dean and J.R.W. Tayloe, NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, Science Application International Corporation, Prepared for the U.S. Nuclear Regulatory Commission, Oak Ridge, TN, 2000.

6. D. Faunce, TMI2-EN-RPT-0002, MCNP Version 6.2 Bias Determination for Low Enrichment Uranium Using the ENDF/B-VIII.0 Cross Section Library, Energy Solutions, 2020.

7. M. Pritchard, NSTS-ES-01, Revision 0, Safe Fuel Mass Limit Criticality Safety Analysis for TMI-2 Decommissioning, Nuclear Safety and Technology Services, 2020.

8. M. Bowen and C. Bennett, NUREG/CR-4604, Statistical Methods for Nuclear Material Management, Pacific Northwest Laboratory, Prepared for the U.S. Nuclear Regulatory

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Commission, Richland, WA, 1988.

9. Nuclear Energy Agency, International Handbook of Evaluated Criticality Safety Benchmark Experiments, Organization for Economic Co-operation and Development.

10. 1992-01-23, GPU, Addition of Borosilicate Glass to the Reactor Vessel

11. NUREG/KM-0001, Supplement 1, "Three Mile Island Accident of 1979 Knowledge Management Digest Recovery and Cleanup."

6.0 ENVIRONMENTAL CONSIDERATION

The NRC staff determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22. Specifically, the following are administrative under 10 CFR 51.22(c)(10):

• Revision, relocation, and deletion of portions of or entire definitions with terms that are only relevant to TMI-2 during PDMS

• Deletion of terms in the POL that are already defined by NRC regulation (i.e., redundant)

• Removal of TSs 3.1.1.1, 3.1.1.2, 3.1.1.3, 3/4.2, 3/4.3, 3/4.4, 5.1 from the POL and relocation to DQAP (6.7.1)

• Removal of certain portions of TSs 6.8, 6.9, and 6.10, and relocation of those portions to the DQAP

• Deletion of TS 6.7.4.b.2 and 6.7.4.b.3 because the TS are already addressed in the ODCM (i.e., redundant)

• Modification of License Condition 2.C.1 to describe the status of the facility

• Deletion of License Conditions 2D, 2E, and 2F because these conditions have been met

• Deletion of the POL Enclosure Statement because it only applies to PDMS

• Changes a definition in the TS Definition Section to reflect facility status and relocates others to the DQAP

• Deletion of TSs 2.0, 3/4.0, 3/4.1, 3/4.2, 3/4.3, 3.3.8, 5.0, 6.0, 6.1, 6.2, 6.3, 6.6, and certain portions of 6.8 because they only apply to PDMS

• Makes editorial changes to the POL including reformatting, repagination, changing margins, changing font, and other minor administrative changes

The following changes meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9):

• Changes to TS 6.11, u se of alternative criteria for HRA

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the license amendment.

7.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Commonwealth of Pennsylvania official was notified of the proposed issuance of the amendment on February 24, 2023 (ML23057A006). The Commonwealth responded on March 9, 2023 (ML23068A468), stating:

Based on our review, we have no comments on the proposed license amendment request or NRCs proposed determination that it involves no significant hazards

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consideration.

8.0 CONCLUSION

Based on the review of TMI-2 Solutions application dated February 21, 2021, as supplemented, the NRC staff finds that the proposed changes are acceptable. Accordingly, the Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:

Jeremy Munson, Senior Nuclear Process Engineer, NMSS/DFM/NARAB Gregory Chapman, Senior Health Physicist, NMSS/DUWP/RDB David Esh, Senior Risk Analyst, NMSS/DUWP/RTAB Raul Hernandez, Safety and Plant Systems Engineer, NRR/DSS/SCPB Andrea Russell, Safety and Plant Systems Engineer, NRR/DSS/STSB, Michael Norris, Senior Emergency Preparedness Specialist, NSIR/DPR/RLB Glenn Tuttle, MC&A Physical Inspection Analyst, NMSS/DFM/MCA Amy M. Snyder, Senior Project Manager, NMSS/DUWP/RDB

Date of Issuance: Revision Issued July 31, 2023

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