ML24334A031

Ns Savannah Safety Evaluation Report for Approval of License Termination Plan

ML24334A031
Person / Time
Site:	NS Savannah
Issue date:	01/30/2025
From:	Tanya Hood Division of Decommissioning, Uranium Recovery and Waste Programs
To:
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ML24334A022	List: ML24334A030 ML24334A031
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Download: ML24334A031 (1)

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1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS RELATED TO AMENDMENT NO. 19 TO FACILITY OPERATING LICENSE NO. NS-1 UNITED STATES MARITIME ADMINISTRATION NUCLEAR SHIP SAVANNAH DOCKET NO. 50-238 January 30, 2025 Contributors Louis Caponi, NMSS Gregory C. Chapman, NMSS Diana Diaz Toro, NMSS Nathan Fuguet, NMSS Linda Gersey, NMSS Tanya E. Hood, NMSS Jack D. Parrott, NMSS Emil Tabakov, NMSS Jean Trefethen, NMSS

2 Table of Contents 1

INTRODUCTION. BACKGROUND, AND REGULATORY EVALUATION.........................................7

1.1 INTRODUCTION

................................................................................................................................7

1.2 BACKGROUND

.................................................................................................................................8

1.3 REGULATORY EVALUATION

..............................................................................................................8 2

SITE CHARACTERIZATION..............................................................................................................12 2.1

SUMMARY

OF CHARACTERIZATION SURVEY METHODS AND RESULTS...............................................13 2.2 HISTORICAL SITE ASSESSMENT.....................................................................................................14 2.3 CHARACTERIZATION ACTIVITIES.....................................................................................................14 2.3.1 Activation Analysis in 2004.................................................................................................15 2.3.2 WPI Characterization in 2005.............................................................................................15 2.3.3 Reactor Internals and Neutron Shield Tank Sampling in 2005...........................................16 2.3.4 Characterization Activities in 2018......................................................................................16 2.3.5 Characterization Activities in 2019......................................................................................18 2.3.6 Survey of Exterior Hall in 2019...........................................................................................20 2.3.7 Sampling Neutron Shield Tank lead in 2021.......................................................................22 2.3.8 Engine Room Survey in 2022.............................................................................................22 2.4 CHARACTERIZATION SURVEY RESULTS..........................................................................................23 2.4.1 Systems..............................................................................................................................23 2.4.2 Structures............................................................................................................................23 2.4.3 NRC Evaluation of Site Characterization............................................................................25 3

REMAINING SITE DISMANTLEMENT ACTIVITIES.........................................................................26 3.1 COMPLETED AND ONGOING D&D ACTIVITIES..................................................................................26 3.2 REMAINING SITE DISMANTLEMENT ACTIVITIES.................................................................................28 3.3 NRC EVALUATION OF REMAINING SITE DISMANTLEMENT ACTIVITIES................................................29 4

PLANS FOR RADIOLOGICAL SITE REMEDIATION.......................................................................29 4.1 NS SAVANNAH REMEDIATION PLANS..............................................................................................30 4.2 NRC EVALUATION OF THE RADIOLOGICAL SITE REMEDIATION PLAN.................................................32 5

FINAL STATUS SURVEY PLAN.......................................................................................................33 5.1 DETERMINATION OF FSS DATA REQUIREMENTS.............................................................................35 5.1.1 Survey Unit Classifications and Area Limitations................................................................35 5.1.2 Reference Coordinate System............................................................................................36 5.1.3 Determining Measurement Locations.................................................................................36 5.1.4 Estimating the Number of Measurements Required...........................................................37 5.1.5 Scanning.............................................................................................................................39 5.1.6 Additional Fixed Measurement Commitments....................................................................40 5.1.7 Surveys for Non-Structural Systems and Components......................................................42 5.1.8 NRC Evaluation of FSS Data Requirements......................................................................42 5.2 RADIONUCLIDES OF CONCERN.......................................................................................................43 5.2.1 Site ROC Determination.....................................................................................................43 5.2.2 Insignificant Contributors....................................................................................................44 5.2.3 NRC Evaluation of ROC Determination..............................................................................44 5.3 SITE RELEASE CRITERIA................................................................................................................45 5.3.1 DCGLs and Dose Factors...................................................................................................45 5.3.2 ALARA Evaluation..............................................................................................................46 5.3.3 DCGLs for Surrogates and Gross Activity Measurements..................................................47 5.3.4 Background and Reference Areas/Measurements.............................................................48 5.3.5 NRC Evaluation of Site Release Criteria............................................................................48

3 5.4 FSS IMPLEMENTATION..................................................................................................................49 5.4.1 Survey Methods..................................................................................................................49 5.4.2 Instrumentation...................................................................................................................50 5.4.3 Quality Assurance Commitments for the FSS....................................................................51 5.4.4 NRC Evaluation of FSS Design and Measurement Quality................................................53 5.5 FSS ASSESSMENT AND REPORT....................................................................................................53 5.5.1 FSS Report.........................................................................................................................55 5.5.2 NRC Evaluation of FSS Assessment and Report...............................................................55 6

COMPLIANCE WITH RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION.......................56 6.1 APPROACH FOR OVERALL DOSE COMPLIANCE................................................................................56 6.1.1 Methods for Evaluating Dose and Establishing DCGLs......................................................57 6.1.2 NRC Evaluation of Approach for Overall Dose Compliance...............................................57 6.2 EXPOSURE SCENARIOS, CRITICAL GROUP, AND PATHWAYS.............................................................58 6.2.1 Exposure Scenarios, Critical Group, and Pathways...........................................................58 6.2.2 Preservation Exposure Scenarios.......................................................................................59 6.2.3 Shipbreaking Exposure Scenarios......................................................................................60 6.2.4 NRC Evaluation of Exposure Scenarios, Critical Group, and Pathways.............................62 6.3 SOURCE TERM..............................................................................................................................64 6.3.1 Tour Guide Source Term....................................................................................................64 6.3.2 Remediation Worker and Component Removal Worker Source Term...............................65 6.3.3 Scrap Yard/Foundry Worker and Leachate Scenarios from NUREG-1640 Source Terms 65 6.3.4 NRC Evaluation of the Source Term Assumptions.............................................................66 6.4 DOSE ASSESSMENT, DCGLS, AND UNCERTAINTY...........................................................................67 6.4.1 Scrap Yard/Foundry Worker Scenario Dose Modeling.......................................................67 6.4.2 Leachate Scenario Dose Modeling.....................................................................................67 6.4.3 Component Removal Worker Scenario Dose Modeling.....................................................67 6.4.4 DCGLs for Radionuclides of Concern.................................................................................70 6.4.5 NRC Evaluation and Independent Analysis of DCGLs and Uncertainty.............................71 6.5 ALTERNATE SCENARIOS: LESS LIKELY BUT PLAUSIBLE....................................................................72 6.5.1 Diver Exposure Scenario....................................................................................................72 6.5.2 Contaminated Seafood Exposure Scenario........................................................................72 6.5.3 NRC Evaluation of Alternate Scenarios: Less Likely but Plausible.....................................74 6.6 NRC EVALUATION OF THE COMPLIANCE WITH RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION CONCLUSIONS.....................................................................................................................75 7

SITE-SPECIFIC COST ESTIMATE....................................................................................................75 7.1 FINANCIAL REQUIREMENTS AND COST ESTIMATE CRITERIA.............................................................76 7.2 NRC EVALUATION OF EVALUATION OF THE UPDATED SITE-SPECIFIC DECOMMISSIONING COST ESTIMATE.................................................................................................................................................76 7.3 NRC EVALUATION OF EVALUATION OF THE DECOMMISSIONING FUNDING PLAN................................77 8

ENVIRONMENTAL CONSIDERATIONS...........................................................................................79 9

PARTIAL SITE RELEASE CONSIDERATIONS................................................................................80 10 STATE CONSULTATION..................................................................................................................80 11 CONCLUSIONS.................................................................................................................................80

4 List of Tables Table 5-2 Investigation Levels (Mark up)....................................................................................40 Table 5-3 Traditional Scanning Coverage Requirements...........................................................40 Table 6-17 (modified) Structural Surface DCGLs corresponding to 15 mrem/y.........................45 Table 5-4 Available Instruments and Associated MDCs (Clean)................................................52 Table 5-5 Initial Evaluation of Survey Results (Background Reference Area Not Used)............54 Table 6-1: Exposure Scenarios Evaluated for Calculation of Surface DCGLs for Preservation and Shipbreaking End-State Conditions.........................................59 Table 6-2 Exposure Scenario Results (mrem/yr)(dpm/cm2)...................................................61 Table 6-3 Surface Contamination Limits (DCGLs).....................................................................70 Table 6-4 DCGLs for all radionuclides........................................................................................71 Table 6-5 Less Likely But Plausible Diver Exposure Scenario DCGLs......................................72 Table 6-6 Concentrations of the Residual Nuclides in the Submerged Containment Volume Compared with NRC Effluent Release Limits in 10 CFR Part 20 Appendix B Surface Contamination Limits (DCGLs)..............................................................73 Table 6-9 Parameters for Remediation Worker on Ship Ingestion Dose Calculations 68 Table 6-10 Parameters for Remediation Worker on Ship Inhalation Dose Calculations 68 Table 6-11 Parameters for Component Removal Worker on Ship External Dose Calculations 69 Table 7-1 Decommissioning Funds Evaluation..........................................................................79 List of Figures Figure 6-1 Receptor and Sources for RESRAD-BUILD Model for Tour Guide Scenario............67

5 List of Acronyms and Abbreviations ADAMS Agencywide Documents Access and Management System ALARA As Low as (is) Reasonably Achievable AMCG Average Member of the Critical Group Bq Becquerel C-14 Carbon-14 CFR Code of Federal Regulations Ci Curie cm Centimeter cm2 Square centimeter Co-60 Cobalt-60 cpm Counts per minute ccpm corrected counts per minute Cs-137 Cesium-137 DCGL Derived Concentration Guideline Limit DCGLEMC DCGL that represents the same dose to an individual for residual radioactivity in a smaller area within a survey unit, by taking into account the area factor DCGLW DCGL for the average residual radioactivity in a survey unit dpm Disintegrations per minute dpm/100 cm2 Disintegrations per minute per 100 square centimeters DQAP Decommissioning Quality Assurance Plan DQO Data Quality Objective EMC Elevated Measurement Comparison FR Federal Register FSAR Final Safety Analysis Report FSS Final Status Survey ft Feet ft2 Square foot H-3 Tritium HPGe High Purity Germanium HSA Historical Site Assessment HTD Hard-to-Detect ISOCS In-Situ Object Counting System LAR License Amendment Request LTP License Termination Plan m

Meter m/s Meters per second m2 Square meters m3 Cubic meters m3/s Cubic meters per second MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual (NUREG-1575)

MDA Minimum Detectable Activity

6 MDC Minimum Detectable Concentration MDCC minimum detectable corrected counts per minute mrem Millirem mrem/yr Millirem per year mSv MilliSievert mSv/yr MilliSievert per year MWth Megawatt thermal Ni-63 Nickel-63 NRC U.S. Nuclear Regulatory Commission pCi/g PicoCuries per gram pCi/L PicoCuries per Liter QA Quality Assurance QAPP Quality Assurance Project Plan QC Quality Control RAI Request for Additional Information RASS Remedial Action Support (In-Process) Surveys RESRAD The RESRAD family of computer codes is a regulatory tool for evaluating radioactively contaminated sites, specifically designed to help determine the allowable RESidual RADioactivity in site cleanup RESRAD-ONSITE A specific code within the RESRAD family of computer codes designed to estimate radiation doses and risks from RESidual RADioactive materials in soils RG Regulatory Guide ROC Radionuclide of Concern ROCs Radionuclides of Concern SER Safety Evaluation Report SOF Sum-of-Fractions Sr-90 Strontium-90 Sv Sievert TEDE Total Effective Dose Equivalent UFSAR Updated Final Safety Analysis Report yr Year

7 1

INTRODUCTION. BACKGROUND, AND REGULATORY EVALUATION 1.1 Introduction By letter dated October 23, 2023 (Agencywide Documents Access and Management System Accession No. ML23298A041), as supplemented by letters dated June 27, 2024, October 16, 2024, and December 19, 2024 (ML24183A271, ML24292A030, ML24358A227, and ML24358A242 respectively), the United States Maritime Administration (MARAD, licensee),

submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC) for the Nuclear Ship Savannah (NS Savannah or NSS) to approve NS Savannahs License Termination Plan (LTP) and add License Condition 2.C.(4) to include LTP requirements and establish criteria for determining when changes to the LTP require prior NRC approval.

The LTP was submitted as a supplement to the NS Savannah final safety analysis report (FSAR) and was accompanied by a requested license amendment that, if approved, in addition to approving the LTP, would add License Condition 2.C.(4), which would establish the criteria for when changes to the LTP require prior NRC approval. The licensee submitted the LAR in accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.90, Application for amendment of license, construction permit, or early site permit.

The NRC accepted the proposed LAR and LTP for review by e-mail dated December 15, 2023 (ML23352A371) and published a notice of consideration of the proposed LAR and a finding of no significant hazards consideration determination in the Federal Register (FR) on March 29, 2024 (89 FR 22199). The notice offered a 30-day comment period and a 60-day period to request a hearing or to petition for leave to intervene. No comments, hearing requests or petitions for leave to intervene were received.

In accordance with Paragraph (a)(9)(iii) of 10 CFR 50.82, Termination of license, the NRC is required to hold a public meeting near the relevant site after the licensee submits an LTP so that NRC staff can discuss the NRC's review of the LTP. This meeting was announced with ads in the Baltimore Sun on April 24, 2024, and May 5, 2024, respectively, and in the Federal Register on April 3, 2024 (89 FR 20375). The meeting was held on May 8, 2024, onboard the NS Savannah in Baltimore, Maryland. The Federal Register notice also included a 60-day opportunity for the public to provide comments on the LTP in accordance with 10 CFR 50.82(a)(9)(iii). An NRC summary of that public meeting was issued on June 7, 2024 (ML24157A373) and includes a summary of questions from the public and the NRC staffs answers. No other public comments were received on the LTP.

By letters dated May 30, 2024 (ML24157A103) and September 16, 2024 (ML24261B855) the NRC sent requests for additional information (RAIs) to the licensee requesting that they provide additional information for the NRCs review of the LTP LAR. As stated above, the licensee responded to those RAIs by letters dated June 27, 2024, and October 16, 2024. The supplements dated June 27, 2024, and October 16, 2024, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the FR on March 29, 2024 (89 FR 22199).

8

1.2 Background

The NS Savannah is the worlds first nuclear-powered merchant ship. Unlike a land-based nuclear plant, the NS Savannah is waterborne, mobile and of unique historic significance. It was first brought to power in 1961 by an 80 MWth pressurized water nuclear reactor. The ship was designed, constructed and operated as a joint research and development project of the U.S.

Department of Commerce, MARAD and the U.S. Atomic Energy Commission (AEC). It was conceived and constructed during the President Eisenhower Administration as a signature element of the Atoms for Peace Program. The AEC issued the initial operating License No.

NS1, Docket No. 50-238 in August 1965.

The NS Savannah was removed from service in 1970. All fuel was removed from the ship in 1971 and taken to the U.S. Department of Energys Savannah River Site in 1972. The NRC issued a Possession-only License in 1976. The Possession-only License No. NS-1 does not authorize operation of the NS Savannah or emplacement or retention of fuel into the reactor vessel. The NS Savannah was listed in the National Register of Historical Places in 1983 and was designated a National Historic Landmark in 1991 as one of the most visible and intact examples of the Atoms for Peace program. The ship is located at Pier 13, Canton Marine Terminal, Baltimore, Maryland. Active decommissioning began in 2018. The ship's reactor pressure vessel was safely removed on November 8, 2022, and transported by rail to Energy Solutions' low-level radioactive waste disposal facility in Clive, Utah.

The NS Savannah LTP provides the details of the plan for characterizing, identifying, and remediating the remaining residual radioactivity at the NS Savannah site to a level that will allow the site to be released for unrestricted use. The NS Savannah LTP also describes how the licensee will confirm the extent and success of remediation through radiological surveys, provide financial assurance to complete decommissioning, and ensure the environmental impacts of the decommissioning activities are within the scope originally envisioned in the associated environmental documents.

1.3 Regulatory Evaluation In accordance with 10 CFR 50.82(a)(9), [a]ll power reactor licensees must submit an application for termination of license. The application for termination of license must be accompanied or preceded by a license termination plan to be submitted for NRC approval. The licensee has not applied for termination of the license at this time.

Under 10 CFR 50.82(a)(9)(i), the LTP must be a supplement to the FSAR, or equivalent. In accordance with 10 CFR 50.82(a)(9)(ii), the LTP must include:

(A)

A site characterization; (B)

Identification of remaining dismantlement activities; (C)

Plans for site remediation; (D)

Detailed plans for the final radiation survey; (E)

A description of the end use of the site, if restricted; (F)

An updated site-specific estimate of remaining decommissioning costs;

9 (G)

A supplement to the environmental report, pursuant to § 51.53, Postconstruction environmental reports, describing any new information or significant environmental change associated with the licensee's proposed termination activities; and (H)

Identification of parts, if any, of the facility or site that were released for use before approval of the LTP.

The approval criteria for the LTP are provided in 10 CFR 50.82(a)(10), which states:

If the license termination plan demonstrates that the remainder of decommissioning activities will be performed in accordance with the regulations in this chapter, will not be inimical to the common defense and security or to the health and safety of the public, and will not have a significant effect on the quality of the environment and after notice to interested persons, the Commission shall approve the plan, by license amendment, subject to such conditions and limitations as it deems appropriate and necessary and authorize implementation of the license termination plan.

Section 1.4, Decommissioning Objectives, of the LTP describes the licensees decommissioning objective as conducting remediation and survey operations such that it can submit a request to the NRC for unrestricted release of the site in accordance with Subpart E,

[Radiological Criteria for License Termination,] of 10 CFR Part 20, [Standards for Protection Against Radiation,] after meeting the requirements of 10 CFR 20.1402, Radiological Criteria for Unrestricted Use.

10 CFR 20.1402 states:

A site will be considered acceptable for unrestricted use if the residual radioactivity that is distinguishable from background radiation results in a [total effective dose equivalent] TEDE to an average member of the critical group that does not exceed 25 [millirem] mrem (0.25

[milliSieverts] mSv) per year [yr], including that from groundwater sources of drinking water, and that the residual radioactivity has been reduced to levels that are as low as reasonably achievable (ALARA). Determination of the levels which are ALARA must take into account consideration of any detriments, such as deaths from transportation accidents, expected to potentially result from decontamination and waste disposal.

The NRC evaluated the LTP to verify that the licensee satisfied the requirements that: 1) the LTP be submitted at least 2 years before termination of the license (50.82(a)(9)(i)); 2) the LTP includes the required parts of an LTP (50.82(a)(9)(ii)); and 3) the plans described in the LTP provide reasonable assurance that the licensee will be able to perform adequate surveys to, if performed consistent with the LTP, demonstrate compliance with the radiological criteria for unrestricted use, as specified in 10 CFR 20.1402.

As part of providing reasonable assurance that the licensee will be able to perform adequate surveys to demonstrate compliance with the unrestricted release criteria in 10 CFR 20.1402, the licensee must also provide reasonable assurance that it will be able to meet the requirements in Subpart F, Surveys and Monitoring, of 10 CFR 20.1501 (a) and (b). Among other things, these regulations require that the licensee has performed or will perform necessary and reasonable surveys to evaluate the potential radiological hazards of the radiation levels and residual radioactivity detected.

10 Footnote 3 of Chapter 1 of the LTP notes that MARAD has adopted a 15 mrem/yr standard for the [NS Savannah]. Within the LTP, direct citations of regulations (i.e., 25 mrem/yr) should be interpreted to mean 15 mrem/yr unless stated otherwise. The NRC staff understands that the licensee has imposed an administrative dose standard of 15 mrem/yr. However, the NRC staff has evaluated the licensees proposed LTP, including its associated dose assessments, against the 25 mrem/yr dose limit in 10 CFR 20.1402. The 15 mrem/yr adopted dose standard has been evaluated as an additional safety margin and provides additional assurance that the licensee will be successful in meeting the established derived concentration guidelines (DCGLs) for license termination.

To perform its review of the NS Savannah LTP, the NRC staff used the guidance in:

Regulatory Guide (RG) 1.179, Standard Format and Contents for License Termination Plans for Nuclear Power Reactors Revision 2, dated July 2019 (ML19128A067);

NUREG-1700, Standard Review Plan for Evaluating Nuclear Power Reactor License Termination Plans Revision 2, dated April 2018 (ML18116A124);

NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), Revision 1, dated August 2000 (ML003761445);

NUREG-1757, Consolidated Decommissioning Guidance, - Volume 1, Revision 2, dated September 2006; Volume 2, Revision 2, dated July 2022; and, Volume 3, Revision 1, dated February 2012 (ML063000243, ML22194A859, and ML12048A683 respectively).

Both RG 1.179 and NUREG-1700 identify NUREG-1575 and NUREG-1757 as guidance for developing site characterization plans, developing remediation plans, developing site-specific DCGLs, demonstrating compliance with the unrestricted release criteria for license termination, and developing final survey methods and plans.

Consistent with 10 CFR 50.82(a)(10), the submitted LTP was accompanied by a proposed license amendment that would approve the LTP. Recognizing that there may be a need to make changes to the LTP following its approval by the NRC, the licensee also included, in the proposed license amendment, a license condition with criteria for when changes to the LTP require NRC approval. In reviewing the license condition, the NRC staff used the guidance and model license condition in Appendix B of NUREG-1700.

The proposed license condition, in the October 23, 2023, submittal, as supplemented by letters dated June 27, 2024, October 16, 2024, and December 19, 2024, would allow the licensee to make certain changes to the approved NS Savannah LTP without prior NRC review or approval.

The proposed License Condition 2.C.(4) would read as follows:

2.C.(4) License Termination Plan (LTP)

MARAD shall implement and maintain in effect all provisions of the License Termination Plan (LTP), dated October 23, 2023, as supplemented by letters dated June 27, 2024, October 16, 2024, and December 19, 2024, as approved in License Amendment No. 19, subject to and as amended by the following stipulations. MARAD may make changes to the LTP without prior NRC approval provided the proposed changes do not meet any of the following criteria:

(a) Require Commission approval pursuant to 10 CFR 50.59.

11 (b) Result in the potential for significant environmental impacts that have not previously been reviewed.

(c) Detract or negate the reasonable assurance that adequate funds will be available for decommissioning.

(d) Decrease a survey unit area classification (i.e., impacted to not impacted; Class 1 to Class 2; Class 2 to Class 3; or Class 1 to Class 3) without providing NRC a minimum 14 calendar day notification before implementing the change in classification.

(e) Increase the derived concentration guideline levels and related minimum detectable concentrations (MDCs) for both scan and fixed measurement methods. If MDCs are increased (relative to what was approved) the licensee should request NRC approval.

(f) Increase the radioactivity level, relative to the applicable derived concentration guideline level, at which an investigation occurs.

(g) Change the statistical test applied to a test other than the Sign Test. Note that the Wilcoxon Rank Sum test will not be used at the NSS.

(h) Increase the approved Type I decision error. Only Scenario A will be used in the FSS of the NSS. Therefore, changing the Type II error when using Scenario B is not applicable and does not require NRC approval.

(i) Change the approach used to demonstrate compliance with the dose criteria (e.g., change from demonstrating compliance using derived concentration levels to demonstrating compliance using a dose assessment that is based on final concentration data).

(j) Change parameter values or pathway dose conversion used to calculate the dose such that the resultant dose is lower than in the approved LTP and if a dose assessment is being used to demonstrate compliance with the dose criteria.

Based on review of the LTP, as documented in this Safety Evaluation Report (SER), the NRC staff has determined that the LTP contains the required information, as described in 10 CFR 50.82(a)(9), in adequate detail to allow for LTP approval. In addition, the NRC staff finds that the criteria in the proposed license condition that would be used to determine if changes to the LTP require NRC approval are equivalent to the list of LTP areas that cannot be changed without NRC approval identified in appendix B of the NUREG-1700.

As described in 10 CFR 50.82(a)(9)(i), the NRC will approve, as appropriate, the LTP as a supplement to the FSAR or equivalent. The NRCs approval of the LTP, as appropriate, is predicated on the site conditions as described in the LTP and the reasonably foreseeable results of the continuing characterization of the site, the implementation of the remaining

12 dismantlement activities and the plans for site remediation, and the LTP becoming a supplement to the FSAR. If the continuing characterization of the site, site dismantlement activities or site remediation activities find, or result in, types or quantities of residual contamination not identified in the LTP, in accordance with 10 CFR 50.59(c), the licensee must evaluate that new information against the methods of evaluating residual contamination described in the LTP to determine if a change is needed to the methods described in the LTP and if so, evaluate that change using the change criteria in the license condition approving the LTP to determine if the change needs prior approval from the NRC.

2 SITE CHARACTERIZATION Chapter 2 of the NS Savannah LTP, Site Characterization, discusses the results of the NS Savannah site characterization activities. In accordance with the requirements of 10 CFR 50.82 (a)(9)(ii)(A) and the guidance of NUREG-1700, the licensee provided Chapter 2, of the NS Savannah LTP, which provides a description of the radiological characterization performed by the licensee.

As stated in Section 2.1 of the LTP, the purpose of site characterization is to ensure that the NS Savannah final status survey (FSS) will be conducted in all areas where contamination existed, remains, or has the potential to exist or remain. The applicable primary objectives of a characterization survey are discussed in Section 2.4.4 of NUREG-1575, as being:

Determine the nature and extent of the contamination Collect data to support evaluation of remedial alternatives and technology Evaluate whether the survey plan can be optimized for using the FSS Provide input to the FSS The Acceptance Criteria in NUREG-1700 for site characterization, applicable to NS Savannah, are:

The LTP identifies all locations, both inside and outside the facility, where radiological spills, disposals, operational activities, or other radiological accidents and or incidents occurred and could have resulted in contamination. This identification should be done on a room-by-room or area-by-area basis as necessary, including equipment, or laydown areas.

The LTP describes, in summary form, the original shutdown, and current radiological and non-radiological status of the site.

The LTP site characterization is sufficiently detailed to allow the NRC to determine the extent and range of radiological contamination of structures, systems (including sewer systems and waste management systems), floor drains, ventilation ducts, piping and embedded piping, rubble, surface water, components, residues, and environment, including maximum and average contamination levels and ambient exposure rate measurements of all relevant areas (structures, and equipment) of the site.

The LTP identifies the survey instruments and supporting quality assurance (QA) practices used in the site characterization program.

The LTP identifies the background levels used during scoping or characterization surveys.

13 The LTP describes in detail the areas and equipment that need further remediation to allow the reviewer to estimate the radiological conditions that will be encountered during remediation of equipment, components, structures, and outdoor areas.

2.1 Summary of Characterization Survey Methods and Results Section 2.1.3, Chronology of Decommissioning Planning and Characterization, of the LTP summarizes the history of the licensees decommissioning planning activities. In 2003, the licensee awarded a decommissioning planning contract to WPI, Inc. WPIs initial decommissioning efforts are described in Section 2.4.1 of the Post Shutdown Decommissioning Activities Report (PSDAR), Revision 1, dated December 11, 2008 (ML083590349). These efforts included the 2004 Reactor Pressure Vessel (RPV) activation analysis described in Section 2.3.1, Activation Analysis in 2004, of the LTP; the activation analysis is documented as report CR-142, Reactor Vessel, Internals and Neutron Shield Tank Characterization and Classification Assessment, dated April 3, 2004. In the first half of 2005, the characterization scoping surveys described in Section 2.3.2, WPI Characterization in 2005, of the LTP were performed. This effort extended throughout the ship and is documented as report CR-038, Radiological and Non-Radiological Spaces Characterization Survey Report, Revision 1, dated February 2, 2006. The effort included radiological and environmental sampling and involved the first entries into the Containment Vessel since 1975.

The LTP continues to describe that one significant finding of the 2005 WPI characterization did not support the 2004 RPV activation analysis, which led to the destructive sampling of the RPV and surrounding Neutron Shield Tank described in Section 2.3.3, RPV, Reactor Internals and

[Neutron Shield Tank] Sampling in 2005, of the LTP. On December 11, 2008, the licensee submitted Revision 1 of their PSDAR (ML083590349) which superseded that 2006 survey report. The licensee provides a summary description of the planning activities associated with decommissioning and license termination. The decommissioning project itself was rescoped, with a near-term focus on returning the NS Savannah to protective storage, with significant administrative and technical upgrades to meet contemporary SAFSTOR criteria.

The full scope of the SAFSTOR effort was not funded; the licensee was able to complete the research and investigations necessary to prepare and complete the Historical Site Assessment (HSA). The HSA was revised in 2023. The licensee completed an Environmental Assessment (EA) and Finding of No Significant Impact (FONSI) dated October 3, 2008 (ML082810182). The 2004 activation analysis documented in CR-142 was the technical basis for the conclusion that the RPV would not meet the waste acceptance criteria for the EnergySolutions radioactive waste disposal facility in Clive, Utah. The competitive pressures relaxed when it was shown, via the 2005 project, documented in CR-038, that the RPV could meet the Clive waste acceptance criteria. Those pressures were then eliminated in 2009 when an additional commercial LLRW repository was opened in Andrews, Texas. In 2019, the licensee received a determination from the Department of Energy (DOE) that NS Savannah waste was eligible for disposal at certain DOE sites. In consultation with DOE, the licensee opted to pursue a commercial alternative for waste disposal.

In 2019, a MARSSIM based survey was conducted on the fully exposed exterior surfaces of the hull, while the ship was on drydock in Philadelphia, PA. This effort is described in Section 2.3.6, Survey of Exterior Hull in 2019, of the LTP and is documented in report CR-143, MARSSIM Survey of the Exterior Hull, dated December 2020. In 2021, the exterior lead shielding on the Neutron Shield Tank (also known as the Primary Shield Tank) was sampled and analyzed to

14 support the release of the material. This effort is described in Section 2.3.7, Sampling [Neutron Shield Tank] lead in 2021, of the LTP and is documented in report CR-144, Primary Shield Water Tank Lead Sample Results, dated April 2022.

Section 2.1.4, Other Considerations Regarding Site Characteristics and Characterization, of the LTP describes a number of unique features and characteristics that distinguish the NS Savannah from typical land-based facilities. Among these are the following:

The site (i.e., the ship) contains no soil; The site contains no surface or groundwater; The site contains no embedded pipe, rubble, or buried or surface paved parking lots; The topside deck gravity drains overboard; Welded hull blanks prevented any auxiliary or secondary system from discharging overboard after operations ceased, and The NS Savannah sewer systems operated as follows:

o Both sampling sinks and the decontamination shower drain in the Cold Water Chemistry Lab (C deck Port) gravity drain to Contaminated Water Tank Starboard TD-6; o

The decontamination shower at frame 125 gravity drained overboard. The hull opening was welded closed in 1976; o

There are no deck drains other than the "shower drain" in the cold chemistry laboratory; o

Sinks and showers gravity drained overboard. The hull opening was welded closed in 1976. These are not contaminated systems; and, o

Toilets drained to the sewage tank in the engine room where it was pumped overboard. The hull opening was welded closed in 1976. This is not a contaminated system.

2.2 Historical Site Assessment Section 2.2, Historical Site Assessment, of the LTP and its subsections provide a summary of the HSA. The HSA for the NS Savannah was completed in 2008. The HSA focused on historical events and routine operational processes that resulted in contamination of the plant systems and rooms within the Radiologically Controlled Area (RCA). The licensees HSA was derived from a review of records maintained to satisfy Paragraph (g)(1) in 10 CFR 50.75, Reporting and recordkeeping for decommissioning planning, which requires the licensee to keep records of spills or other unusual occurrences involving the spread of contamination in and around the facility, equipment, or site, and telephone and in-person interviews with former individuals involved in nuclear operations at the NS Savannah. Plant records included routine radioactive effluent release reports, non-routine reports submitted to the NRC under provisions of the technical specifications, 10 CFR Part 20, or 10 CFR Part 50; plant incident reports; corrective action reports; and findings documented in accordance with other assessment processes such as the Decommissioning Quality Assurance Plan (DQAP) and oversight activities. Section 2.2.3, Results, of the LTP provides a summary of the events identified during the HSA including worker contamination events and several primary system piping leaks. The licensee did not identify any significant radiological events.

2.3 Characterization Activities

15 Section 2.3, Characterization Activities, of the LTP, describes the licensees site characterization efforts, performed between 2003-2022, and includes eight detailed characterization surveys throughout the ship to support decommissioning and license termination planning and the acquisition of contract services to perform the related dismantlement, waste disposal, remediation, and survey work. The WPI characterization done in 2005 serves as scoping information for the 2018 and 2019 characterization surveys, while the activation analysis completed in 2004, the reactor internals and Neutron Shield Tank sampling in 2005, and the Neutron Shield Tank sampling in 2021 provide information on waste classification and disposal.

2.3.1 Activation Analysis in 2004 Section 2.3.1 of the LTP summarizes the activation analysis program completed in 2004 and documented in CR-142. The objective of this program was the analytical determination of nuclide activation levels in the NS Savannah RPV, reactor internals and Neutron Shield Tank. In its letter dated May 30, 2024, the NRC staff asked the licensee to clarify the purpose of the activation analysis in 2004, the RPV, Reactor Internals and Neutron Shield Tank sampling in 2005, and the Neutron Shield Tank lead sampling in 2021 and explain if they were used to inform site characterization and development of radionuclides of concern (ROCs) or solely used for the purpose of waste classification. In its RAI response dated June 27, 2024, the licensee clarified in response RAI 2-1a that, the purpose of 2004 activation analysis and 2005 sampling and analysis activities was to determine the waste classification of the RPV. The NRC staff notes that because the RPV has been removed from the ship, the discussion of this characterization step is irrelevant to the ships status at the time of license termination.

2.3.2 WPI Characterization in 2005 Section 2.3.2 of the LTP summarizes the radiological surveys and sampling in report CR-038.

The 2005 NS Savannah Characterization Project was intended to provide the licensee with a profile of radiological contaminants and environmental hazards on the ship in radiological spaces, but as this characterization activity aged, the licensee determined that it effectively devolved into a scoping characterization. In its letter dated May 30, 2024, the NRC staff asked the licensee to clarify the purpose of the WPI characterization in 2005 and describe to what extent it was used to inform the 2018 and 2019 characterization efforts. In its RAI response dated June 27, 2024, the licensee clarified in response RAI 2-1b that the intent of the WPI 2005 characterization program was to provide a basis for which the government could estimate the cost of performing the decommissioning.

The LTP continues to describe that only those locations and equipment/structures that were expected to be radioactive were surveyed in-depth to determine the extent and types of radioactive materials present. The remaining areas (principally aft of the engine room, forward of the reactor compartment, and in the mid-ship-house and public areas) were surveyed less rigorously than radiological areas but in sufficient detail to confirm that no radioactive materials reside in those locations. In addition, WPI took several smears and samples in non-radiological spaces to facilitate future analyses. Included were 1,423 smears, 26 paint samples, 14 metal samples, six core bores in the concrete secondary containment, 10 crud samples from the primary system, four water samples from the primary side of the steam generators, one water sample from the essentially empty Neutron Shield Tank, and 11 air samples.

In-situ gamma spectroscopy was also performed in 16 locations using the Berkeley Nucleonics SAM 935 portable surveillance and measurement system, which uses a 3-inch by 3-inch NaI

16 detector to provide isotopic identification. Each portable instrument was checked daily for proper background. This background value was established when the instrument was first put into service on this project. A source count value using an appropriate check source was established initially for the portable instruments. From this initial count, a +/-20% range was established for each instrument. On a daily basis or more frequently if appropriate, the appropriate check source was counted with each portable instrument. The daily source count was entered on the Instrument Source Check Log for each instrument and verified to be within this +/-20% range.

Table 2-2, 2005 Characterization Project Instrumentation List, in the LTP presents the portable instrumentation used for the project.

The non-radiological areas evaluated are summarized in Table 2-3, 2005 Non-Radiological Area Summary in the LTP. Table 2-4, 2005 Radiological Area Summary, in the LTP provides a summary of the radiological conditions found during the evaluation of radiological areas, excluding containment. Table 2-5, 2005 Reactor Compartment Radiological Summary, in the LTP provides a summary of radiological conditions found during the evaluation of Reactor Compartment areas outside of the Containment Vessel. Table 2-6, 2005 [Containment Vessel]

Radiological Summary, in the LTP provides a summary of the radiological conditions of the Reactor Compartment outside the Containment Vessel. Each of these tables includes the maximum dose rate in each deck/compartment, and any contamination identified. Both the Hold Deck (i.e., Horseshoe Tank Top) and Cargo Hold Number 4 (aft) were subsequently designated as Radiological Areas, due to the elevated dose rates.

2.3.3 Reactor Internals and Neutron Shield Tank Sampling in 2005 Section 2.3.3 of the LTP summarizes the 2005 project to determine the curie content and isotopic inventory of the RPV, Reactor Internals, and Neutron Shield Tank by extracting metal samples at selected locations. In its letter dated May 30, 2024, the NRC staff asked the licensee to provide assurance that there was a complete characterization determining the extent and range of radioactive contamination anticipated to remain on site, including systems and components. In its RAI response dated June 27, 2024, the licensee stated in response RAI 2-4a, that only the secondary side portions of the Steam Generators, the Neutron Shield Tank outer shell and the Pressurizer shell with no internals will remain at the time of license termination. The licensee elaborated, in response to RAI 2-4b, that previous characterization of these components in the 2005 sampling event did not identify any activation of these materials but that it has developed a work package to draw metallurgical samples that will be evaluated by an independent contractor. The licensee stated that the evaluation will be provided to the NRC when complete.

2.3.4 Characterization Activities in 2018 Section 2.3.4, Characterization Activities in 2018, of the LTP, describes the radiological characterization conducted in specific areas outside the reactor compartment and containment vessel by Radiation Safety and Control Services Inc. (RSCS) during August and September 2018. The intention of this work was not to repeat previous characterization efforts, but to fill in the gaps and expand the characterization data in certain areas. The areas surveyed are presented in Table 2-8, 2018 Areas Surveyed in the LTP. Table 2-8 identifies 25 survey units, their descriptions, and the locations/system.

The LTP continues to describe that the primary focus of radiological characterization was a MARSSIM based approach using scans, static measurements, smears and dose rate measurements in 25 identified survey units. In addition, several systems/items were also

17 included in this characterization program. Computer-aided design drawings of the survey units were created specifically for documentation - including floors, walls, and ceilings. Seventeen static/smear locations were designated in each survey unit using Visual Sample Plan (VSP) software. A total of ten tritium samples were part of this characterization and survey units with the highest potential for tritium were selected for these samples. The MARSSIM survey class included specific instructions for collection of measurements and data logging, figures with locations, and documentation for recording survey activities/results. These surveys were designed to collect the specified radiological samples and analysis for these areas (scan, static, smear, dose rate, sample).

For the survey design, the Co-60, H-3, and Am-241 screening values, taken from Table 5.19 of NUREG/CR-5512, Residual Radioactive Contamination from Decommissioning, Volume 3, Parameter Analysis, dated October 1999 (ML082460902), were used as investigation levels.

The investigation levels for beta and gamma scans were 500 counts per minute (cpm) and 5,000 cpm respectively.

A characterization survey package was created for each survey unit, which included specific instructions for collection of measurements and data logging, figures with locations and documentation for recording survey activities/results. Radiological surveys were performed with the following instrumentation:

Thermo RadEye SX with Ludlum Model 43-89 detector (alpha/beta directs and beta scans);

Thermo RadEye SX with Ludlum Model 44-10 (gamma scans);

Ludlum Model 3 with Ludlum Model 44-9 (beta scans);

Bicron MicroRem (dose rate); and, Ludlum Model 19 (dose rate).

Instruments were properly calibrated, and beta efficiencies were determined with Tc-99, alpha efficiencies with Th-230, and gamma calibrations with Cs-137. Operational checks were performed each day prior to use. With a 3-minute count time and 1-minute background, direct measurement MDCs were in the range of 10 to 39 dpm/100cm2 alpha and 98 to 338 dpm/100cm2 beta. Measurements were collected with the detector approximately 0.5 inches from the surface.

In-Situ Object Counting System (ISOCS) measurements were performed with a 2x2 inch stabilized and ISOCS characterized sodium iodide detector with a lead collimator that was portable for access to the various locations within the ship. The ISOCS sensitivity varied by count time, the object being counted, and the distance from the object.

Smears were counted at the RSCS laboratory on a Tennelec gas proportional counting system.

The minimum detectable activity for alpha smear measurements was approximately 16 dpm/100cm2 and approximately 43 dpm/100cm2 for beta smears. In accordance with the DQAP and the Radiation Protection Program (RPP), GEL Laboratories, LLC (GEL) analyzed tritium smears (method GL-RAD-A-002) and also counted 48 smears for QA by gas proportional counting (method GL-RAD-A-001).

The radiological data generated was intended only to provide guidance for future decontamination and remediation activities for these areas. Use of properly calibrated instruments with operational checks and duplicate measurements as part of the survey process

18 provided data quality indicators. The data quality was acceptable with no deviations from measurement protocols and reasonable agreement with duplicate measurements.

The following was performed as described by the LTP during this characterization program:

396 alpha/beta direct measurements; 412 alpha/beta removable measurements; 10 tritium measurements; 403 dose rate measurements; Beta scans; Gamma scans; and, 69 ISOCS measurements.

Table 2.9, 2018 Maximum Results by Measurement Type, in the LTP provides a summary of each type of measurement, the maximum results for each type of survey conducted, the survey unit, and description of the survey unit. Tables 2-10, [NS Savannah] 2018 Characterization Summary Table A, in the LTP and Table 2-11, [NS Savannah] 2018 Characterization Summary Table B, in the LTP summarize the results by survey unit.

2.3.5 Characterization Activities in 2019 Section 2.3.5, Characterization Activities in 2019, of the LTP summarizes the radiological and non-radiological surveys on the ship in radiological spaces performed in 2019, by RSCS. This scope of work was to perform a radiological and environmental hazard characterization for the reactor compartment and containment vessel. Similar to the 2018 characterization effort, the intention of this work was not to repeat previous characterization efforts, but to fill in the gaps and expand the characterization data in certain areas.

Radiological surveys were performed with the following instrumentation:

Ludlum Telepole; Thermo RadEye SX with Ludlum Model 43-93 detector (alpha/beta directs and beta scans);

Thermo RadEye SX with Ludlum Model 44-10 (gamma scans);

Ludlum Model 3 with Ludlum Model 44-9 (beta scans);

Bicron MicroRem (dose rate);

Ludlum Model 19 (dose rate);

AMP-100 dose rate probe; Ludlum 3030E with a 43-10-1 probe above a fixed position smear counter; and, NaIs and OSPREY ISOCS Characterized.

The LTP continues to describe that the instruments were properly calibrated, and beta efficiencies were determined with Tc-99, alpha efficiencies with Th-230, and gamma calibrations with Cs-137. Source check responses were established for each instrument with plus and minus two standard deviation values determined. Operational checks were performed each day prior to use. Control charts for background and daily source checks were established for the counting instruments such as the RadEye with 43-83 alpha-beta probe and the Ludlum 3030 with the 43-10-1 gross alpha/beta probe.

19 For the survey design, the Co-60, H-3, and Am-241 screening values, taken from NUREG/CR-5512, Vol. 3, Table 5.19, were used as investigation levels. Equation 2-1 in the LTP defines the calculation used for minimum detectable activity. Baseline radiation and contamination surveys were performed in the reactor compartment and containment vessel between April 5 and April 30, 2019. The surveys were conducted to evaluate area conditions for end-state and decommissioning planning. MARSSIM scans were not performed since component removals or remediations during decommissioning were deemed likely to change the locations of surface contamination.

The surveys consisted of:

General Area, contact, and 30 cm dose rates (not background corrected);

100cm2 removable contamination smears and gross wipes/smears; Biased scans with alpha-beta or beta-gamma detectors; and, Fixed position counts with alpha-beta detectors.

The quantity of smear samples and measurements for the baseline surveys are summarized in Table 2-12, 2019 Baseline Survey Measurement Summary (Quantity Collected), in the LTP. In addition to the 273 100cm2 smears documented on the surveys, several dozen large area smears were obtained to get sufficient general area surface contamination activity for a reliable radionuclide mix to be generated.

Scans and fixed position measurements are reported by the LTP in corrected counts per minute (ccpm) with background levels in the areas ranging from 0-0.5 cpm alpha and 100- 20,000 cpm beta. The highest alpha and beta backgrounds reported in each area in Table 2-12 were used along with Equation 1 to calculate the minimum detectable corrected counts per minute (MDCC) and minimum detectable activity (MDAs) for fixed position readings that used a 1-minute background and a 3-minute count time. Comparison of the MDAs to the screening DCGLs in Table 2-13, Fixed Position Alpha () Beta () Sensitivity in Highest Background per Survey Area, in the LTP demonstrates that adequate sensitivity was maintained even in the highest background locations with worst case MDAs of 49% of the Co-60 [7,100 dpm/100cm2 DCGL]

and 12% of the Cs-137 [28,000 dpm/100cm2 DCGL].

The system characterization was focused on obtaining representative samples of interior contaminants. This included smears as well as samples of any sludges, liquids, or other materials present. It also focused on obtaining interior beta and gamma dose rates to supplement dose to curie estimates of the source terms present. System access and sampling was performed in accordance with a detailed Work Order and daily Job Hazards Analysis briefs.

Area surveys, system surveys, and sampling were performed using project-specific procedures for the instruments used and sampling performed.

On site gamma spectroscopy measurements were performed with a 2x2 stabilized and characterized sodium iodide detector with a lead collimator that was also portable for access to the various locations within the ship. The geometry composer software was used to create a smear composite, paint chip sample, as well as liquid and sludge sample container geometries for on-site analysis.

In accordance with the DQAP and the RPP, GEL also analyzed baseline survey smears with positive activity for quality assurance by gas proportional counting and tritium smears. System smear sample composites were also analyzed by gamma spectroscopy at the same laboratory.

20 These results were reviewed and five of the smear composites were selected for hard-to-detect analysis to provide C-14, Ni-63, Sr-90 and Tc-99 results. These ROCs were obtained during characterization planning. The bolded and italicized five ID cells in Table 2-14, Off-site Laboratory Smear Composite Gamma Spectroscopy Results, in the LTP were selected for hard-to-detect analysis. The required MDA for gamma spectrometry was set at 25 pCi/filter for Cs-137.

Results of characterization surveys on external surfaces of system components and decks show that there is very little removable and total activity. The results of internal system component surveys only established removable levels and not total activity or whether radioactivity has penetrated the material.

Table 2-15, Steam Generator Composite Smear Results Compared to Screening DCGLs, in the LTP presents the off-site laboratory results for the steam generator composite with the hard-to detect radionuclides. The required MDA for H-3 analysis was 150 pCi/filter. The required MDA for C-14 analysis was 10 pCi/filter. The required MDA for Ni-63 analysis was 20 pCi/filter.

The required MDA for Tc-99 analysis was 10 pCi/filter. The required MDA for Sr-90 analysis was 2 pCi/filter.

Table 2-16, Pressurizer Composite Smear Results Compared to Screening DCGLs, in the LTP presents the off-site laboratory results for the Pressurizer composite with the hard-to-detect radionuclides. Table 2-17, Containment Drain Tank Composite Smear Results Compared to Screening DCGLs, in the LTP presents the off-site laboratory results for the Containment Drain Tank (PD-T4) composite with the hard-to-detect radionuclides. Table 2-18, [Reactor Compartment] Exhaust Ventilation Composite Smear Results Compared to Screening DGCGLs, in the LTP presents the off-site laboratory results for the Reactor Compartment Exhaust Ventilation composite with the hard-to-detect radionuclides. Table 2-19, Primary Loop

[Reactor Compartment] IX Piping Composite Smear Results Compared to Screening DCGLs, in the LTP presents the off-site laboratory results for the Primary Purification Reactor Compartment IX piping composite with the hard-to-detect radionuclides.

The composite activities in Tables 2-15 through Table 2-19 of the LTP were compared to the 90th percentile building screening values from Table 5.19 in NUREG/CR-5512, Volume 3. There is no value for Ag-108m in the screening table; therefore, the value for Ag-110m was used as a substitute for the calculations.

The LTP concludes in describing that the results of the RPV, Internals and Neutron Shield Tank sampling in 2005 were also re-evaluated in 2019. The 2005 activated metal sample results were decayed to 2019. The decayed activity concentrations were converted from Ci/g to Ci/cc by multiplying by the average density of steel at 8 g/cc to determine the waste classification in accordance with 10 CFR Section 61.55, Waste classification. Table 2-20, Component Activity and Classification Summary, in the LTP presents the total activity concentrations and classification of the activated components. Five components were identified as Class A waste, and two components were classified as Class B waste.

2.3.6 Survey of Exterior Hall in 2019 Section 2.3.6 of the LTP describes a MARSSIM based radiological survey performed by RSCS on the NS Savannah exterior hull in September and October of 2019 while the ship was on drydock in Philadelphia, PA. This survey was performed while the ship was on drydock because this was the only time that the exterior hull would be readily accessible to perform these

21 surveys. This survey was performed following the methodology of a FSS, as described in the decommissioning activities guided by MARSSIM.

The LTP continues to explain that as part of the drydock maintenance activities, the underwater portion of the hull was stripped of paint. These surveys were performed after paint stripping (on the bare hull metal) to document the radiological conditions of the hull structural materials directly where available.

The survey design was a MARSSIM based approach using scans, static measurements, smears, and dose rate measurements in nine identified survey units. VSP software was used to create maps and to randomly select sample locations within each survey unit.

The hull was designated as a MARSSIM Class 3 area. The exterior hull of the ship contained a total of nine survey units. Table 2-21, Hull Survey Units, in the LTP provides the FSS Unit Number and a description of the area. Figures 2-1 and 2-2 depict the starboard side and the port side survey units, respectively.

The screening DCGL for Co-60 (most restrictive) of 7,100 dpm /100cm2 for beta (NUREG 1757, Volume 2, Appendix H, Table H.1) was used for design of these surveys. Alpha emitting radionuclides were not included in the screening DCGLs based upon the results of the recent characterization of the Containment Vessel and Reactor Compartment showing no significant alpha were present; therefore, the planned approach did not include the scanning or static measurements for alpha on the hull.

These surveys were designed to collect the specified radiological samples and analysis for these areas (scan, static, smear, dose rate) as laid out in MARSSIM. For Class 3 survey areas, surface scan surveys have judgmental coverage, and given the degree of difficulty to access the survey areas, scan surveys are planned to be performed in the vicinity of the static locations.

The following measurements were performed during this MARSSIM Type Survey:

180 beta direct measurements; 180 beta removable measurements; 180 gamma dose rate measurements; and, Beta scan surveys.

Smears were counted at the RSCS laboratory on a low background gas proportional counting system.

Radiological surveys were performed with the following instrumentation:

Thermo RadEye SX with Ludlum Model 43-89 dual-phosphor scintillation detector (beta scans);

Thermo RadEye SX with 43-93 dual-phosphor scintillation detector (direct static beta measurements and beta scans); and, Bicron MicroRem (dose rate measurements).

The surveys were performed with calibrated instrumentation and operationally checked daily.

Beta efficiencies were conservatively determined with Tc-99 and gamma efficiencies with Cs-137.

22 A MARSSIM survey package was created for each survey unit, which included specific instructions for collection of measurements and data logging, figures with locations, and documentation for recording survey activities/results. A summary of the survey results is provided in Table 2-22, Descriptive Statistics, Beta Static Measurement Data, Background Corrected, in the LTP and Table 2-23, [NS Savannah] Hull MARSSIM Survey Summary, in the LTP.

The LTP section concludes in explaining that the MARSSIM survey of the exterior hull demonstrates that:

No unexpected results or trends are evident in the data; The sampling and survey results demonstrate that residual radioactivity in the survey areas are indistinguishable from background levels; The data quality meets the necessary requirements and is deemed to be acceptable for its intended purpose; The amount of data collected from each survey unit is adequate to provide the required statistical confidence needed to decide that the DCGLs are met; and, All measurements were below the screening DCGL.

2.3.7 Sampling Neutron Shield Tank lead in 2021 Section 2.3.7 of the LTP describes the characterization of the Neutron Shield Tank conducted in October 2021, to prepare for the removal and release of the Neutron Shield Tank lead shielding.

Lead samples were collected by drilling six 1/4 inch diameter holes into the lead shielding at approximately 60-degree intervals. These holes were drilled below the shield tank cooling coils in the region of the highest neutron flux. Figure 2-3, Neutron Shield Tank, shows the active core region by the black rectangle in the center of the figure. Two samples were collected at each location, an inner and outer sample. The purpose of the sampling was to verify whether any neutron activation occurred in the lead. In accordance with the DQAP and the RPP, samples were sent to GEL for gamma spectrometry and analysis of hard-to-detect radionuclides. Only one sample had a positive result at 0.15 pCi/g Cs-137. This result was attributed to cross contamination.

2.3.8 Engine Room Survey in 2022 Section 2.3.8, Engine Room Survey in 2022, of the LTP describes the characterization of the Engine Room. The licensee determined that a more detailed in-depth survey of the secondary systems contained in the engine room would be prudent to support license termination planning, especially with respect to the shipbreaking end-state condition. The effort included removal of asbestos-containing insulation materials (ACM) as a prerequisite activity. With the ACM removed, a survey package was developed and executed in 2022.

The survey package identified eighteen survey points of the internal surfaces of the Engine Room steam, condensate, and feedwater system components for measurements. For all but one survey point, either two or four locations for measurements were specified; for survey point, number eighteen, eight locations for measurements were specified. The technicians were required to collect a three-minute Total Surface Activity measurement at the survey location with a Ludlum model 43-93 detector, collect a 100 cm2 smear at the Total Surface Activity location and perform a beta scan survey of 100% of the accessible system internals. The technicians

23 were also required to collect a three-minute local area instrument background (beta and alpha) with the model 43-93 detector at or in the general area of the survey location. The MDCs ranged from 311 to 377 dpm/100 cm2. Six smears were sent offsite for gamma isotopic analysis. Only one smear had positive Co-60 activity of 23 dpm/filter. This smear was from the Main Steam Moisture Separator. Based upon this analysis, the Main Steam system has been classified as impacted.

2.4 Characterization Survey Results Section 2.4, Initial Classifications, of the LTP and its subsections, describe the impacted /non-impacted systems and the classifications for structures. The initial classifications were obtained from the preliminary classifications presented in the HSA, which were determined from operational history and records of spills or other unusual occurrences involving the spread of contamination in and around the facility, equipment, or site. Those preliminary classifications were based upon the screening values for residual surface radioactivity for building surfaces presented in NUREG-1757, Volume 2, Table H-1. In addition, the sites operational RPP, along with the 2004 characterization scoping surveys completed by WPI, in report CR-038, provide valuable historical data and continues to provide input regarding site radiological conditions.

During development of the LTP, the licensee completed reviews of the radiological survey data (routine and non-routine) over the ships operating history since 2008 and the 2018 and 2019 characterization survey results compiled for decommissioning.

2.4.1 Systems Section 2.4.1, Systems, of the LTP indicates which system is impacted or non-impacted.

During development of the LTP, an extensive review of systems was conducted by the licensee to determine those systems that contain radioactive materials or in which radioactive material was detected at some time during the operating history of the plant. Table 2-24, Initial Classification of Systems, in the LTP provides a listing of plant systems and their status relative to the potential for radioactivity (impacted or non-impacted). Seventeen systems are impacted, seven systems are non-impacted. The assessment considers the internal portions of the systems. Systems that might be assessed as non-impacted and are located in contaminated areas may themselves be externally contaminated and may be considered for remediation or disposal as radioactive waste.

2.4.2 Structures Section 2.4.2, Structures, of the LTP describes the classification of structures. Classification is the process by which an area or survey unit is described according to its radiological characteristics. The significance of classification is that this process determines the FSS survey design and the procedures used to develop this design. In classifying areas, those that have no reasonable potential for residual contamination are classified as non-impacted areas. These areas have no radiological impact from site operations and are typically identified early in decommissioning. Areas with some potential for residual contamination are classified as impacted areas. Impacted areas are further divided into one of three classifications as defined by MARSSIM:

Class 1 areas: Areas that have, or had prior to remediation, a potential for radioactive contamination (based on site operating history) or known contamination (based on previous radiological surveys) above the DCGL. Examples of Class 1 areas include: 1) site areas previously subjected to remedial actions, 2) locations where leaks or

24 unplanned releases are known to have occurred, 3) waste storage sites, and 4) areas with contaminants in discrete solid pieces of material with high specific activity. Areas containing contamination in excess of the DCGL prior to remediation would generally be classified as Class 1 areas unless ample evidence exists to show that a lower classification is justified; Class 2 areas: These areas have, or had prior to remediation, a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGL. To justify changing the classification from Class 1 to Class 2, there should be measurement data that provides a high degree of confidence that no individual measurement would exceed the DCGLW. Other justifications for reclassifying an area as Class 2 may be appropriate based on sites-specific considerations. Examples of areas that might be classified as Class 2 for the final status survey include: 1) locations where radioactive materials were present in an unsealed form (e.g., process facilities), 2) potentially contaminated transport routes, 3) areas downwind from stack release points, 4) upper bulkheads and overheads of some buildings or rooms subjected to airborne radioactivity,

5) areas where low concentrations of radioactive materials were handled, and 6) areas on the perimeter of former contamination control areas; and, Class 3 areas: Any impacted areas that are not expected to contain any residual radioactivity or are expected to contain levels of residual radioactivity at a small fraction of the DCGL, based on site operating history and previous radiological surveys.

Examples of areas that might be classified as Class 3 include buffer zones around Class 1 or Class 2 areas, and areas with very low potential for residual contamination but insufficient information to justify a non-impacted classification.

The LTP continues to explain that Class 1 areas have the greatest potential for contamination and therefore receive the highest degree of survey effort for the FSS, followed by Class 2 areas, and then by Class 3 areas. Non-impacted areas do not require any level of survey coverage because they have no potential for residual contamination. As a survey progresses, reevaluation of classifications may be necessary based on newly acquired survey data. The FSS plan includes a process by which measurements that approach pre-defined action levels (fractions of the DCGLs) are investigated to see if re-classification of an area(s) is necessary.

In its letter dated May 30, 2024, the NRC staff asked the licensee to explain the inconsistencies between the information provided in Table 2-8 and the information provided in Tables 2-24 and 2-25. In its RAI response dated June 27, 2024, the licensee updated Table 2-25 in its RAI 2-2 response to reflect that some ship areas have been consolidated and others have been renamed to more familiar descriptions since 2018. The survey results will be notated in the FSS packages with the survey unit number shown in Table 2-25, Initial Classification of Structures or Rooms, in the LTP. The preliminary survey unit numbers assigned in the 2018 characterization effort will not be used.

The LTP continues to describe that based upon the nature and extent of contamination detected and the areas to be surveyed, some of the survey unit decks and bulkheads below 2 meters high will be treated as either a Class 1 or Class 2 survey unit. Depending on the area, the bulkhead above 2 meters and the overhead will be treated as either a Class 2 or Class 3 survey unit.

In its letter dated May 30, 2024, the NRC staff asked the licensee about the impact of volumetric contamination. In its RAI response dated June 27, 2024, the licensee stated in response

25 RAI 2-4, that there is no volumetric contamination in the materials that will remain on the NS Savannah. Additional samples were requested by the NRC staff for verification and the licensee has assigned the development of the survey plan and evaluation to an independent contractor who will provide the evaluation to the NRC when complete.

2.4.3 NRC Evaluation of Site Characterization The NRC staff evaluated the site characterization and HSA in the NS Savannah LTP in accordance with Section 2.2, Site Characterization, of NUREG-1700. As described in NUREG-1700, the purposes of the NRCs review are: (1) to ensure that the site characterization presented in the LTP is complete; and, (2) to verify that the licensee obtained the data using sufficiently sensitive instruments and proper QA procedures to obtain reliable data that are relevant to determining whether the site will meet the decommissioning limits if characterization data is used as final survey data. The acceptance criteria for Section 2.2 of the NUREG-1700 states that the LTP should: (1) identify all locations where activities (including spills) could have resulted in contamination; (2) summarize the status of the site; (3) be sufficiently detailed to allow a reader to determine the contamination levels; (4) identify survey instruments and practices; (5) identify background radiation levels; and, (6) describe areas and equipment that need further remediation.

In addition, the NRC staff evaluated the site characterization information using the guidance contained in NUREG-1757 and MARSSIM (NUREG-1575), which describes ways to meet the objectives of providing an adequate site characterization, as required by 10 CFR 50.82(a)(9)(ii)(A), as well as ways for the site characterization information to provide a necessary and reasonable evaluation of the site characterization and recordkeeping requirements described in 10 CFR 20.1501(a) and (b).

The LTP summarized the past characterization efforts and the HSA. The HSA identified those known locations where atmospheric releases, unplanned liquid releases, facility contamination, and release of radioactive material occurred prior to the LTP submittal and could have resulted in contamination within or outside of the facility. In its letter dated May 30, 2024, the NRC staff asked the licensee to clarify if any inaccessible areas remain where characterization still needs to occur and to provide plans for future characterization. In its RAI response dated June 27, 2024, the licensee stated in response RAI 2-3, that there are no inaccessible areas on the [NS Savannah] and we have completed characterization. As such, the NRC staff finds that there were reasonable efforts made to characterize and evaluate the contamination levels in accessible portions of the site.

The licensee evaluated potential ROCs in the characterization studies, including Ag-108m, C-14, Co-60, Cs-137, Ni-63, Sr-90, Tc-99, H-3, and Fe-55, Am-241, and Pu-239/240. Hard-to-detect radionuclides included C-14, Ni-63, Sr-90, and Tc-99. The sampling results from the characterization studies indicated that only Cs-137 and Co-60 are the Radionuclides of Concern for the NS Savannah final end-state and the remaining radionuclide are Insignificant Dose Contributors. This determination is found acceptable by the NRC staff.

The licensee has sufficiently detailed the status of the NS Savannah to preliminarily determine the extent and range of radiological contamination in the systems and structures. The characterization data supplied is sufficient to preliminarily determine areas approaching or exceeding the release criteria when compared to preliminary interim screening values. The NRC staff finds that the characterization was adequate to identify areas where remediation will be performed, and that the licensee had estimated the radioactive waste streams accordingly.

26 The LTP describes the systems and structures that are considered impacted and further divided into one of the three classification areas as defined by MARSSIM. The NRC staff finds the licensees justification for non-impacted areas adequate, based on the results of the characterization studies and historical data obtained from the HSA.

In sum, the NRC staff finds that the NS Savannah LTP meets the acceptance criteria as delineated in Section 2.2 of NUREG-1700. In addition, the NRC staff evaluated the licensees site characterization survey practices and instruments. Based on this review, the NRC staff determined that the licensee has met the 10 CFR 50.82(a)(9)(ii)(A) requirement that the LTP to include [a] site characterization as well as the requirements of 10 CFR 20.1501(a) and (b).

3 REMAINING SITE DISMANTLEMENT ACTIVITIES In accordance with the requirements of 10 CFR 50.82(a)(9)(ii)(B), the LTP must identify the remaining major dismantlement and decontamination (D&D) activities for the decommissioning of the site at the time of LTP submittal. The licensee followed the guidance of RG 1.179 and NUREG-1700 to provide that information in Chapter 3, Identification of Remaining Site Dismantlement Activities, of the LTP. Those activities can be undertaken before approval of the LTP pursuant to 10 CFR 50.82(a)(5) and (6) and the current 10 CFR Part 50 license for NS Savannah (Possession-Only License No. NS-1), and will be consistent with the PSDAR.

The guidance in RG 1.179 and NUREG-1700 describes that the LTP should include:

A discussion of the remaining D&D tasks, decontamination techniques, and projected schedules for use in planning further decommissioning activities and for NRC to identify any inspection or technical resources needed during the remaining dismantlement activities.

A description of the proposed control mechanisms to ensure that areas are not re-contaminated.

Occupational exposure estimates and a characterization of the type and quantity of radioactive waste produced.

A description of how the remaining activities are evaluated for unreviewed safety questions or against the facilitys licensing requirements.

3.1 Completed and Ongoing D&D Activities As stated in the LTP, at the time of submittal, all dismantlement was essentially complete.

Shipment of LLRW will continue until complete.

Section 3.1.1, Dismantlement Scope and Planned Final Ship Configuration, of the LTP includes a discussion of the dismantlement scope and planned final ship configuration; the completed and ongoing decommissioning activities; a characterization of the radiological impacts of these remaining decommissioning activities; and radioactive waste characterization and projections.

27 The LTP explains that the scope of dismantlement for the ship described in the PSDAR is based on several fundamental assumptions, which are supported by the initial characterization efforts described in Chapter 2 of the LTP. Among the assumptions are:

a) The ship itself is not dismantled as part of DECON; b) Existing accesses are utilized to support dismantlement of systems and components;

and, c)

Major structures will not be dismantled.

These assumptions are based, in part, on National Historic Preservation Act (NHPA) requirements and satisfactory FSSs. Among the initial structures planned to be retained are the Containment Vessel and its foundation, and the Secondary Shield. Primary ship structures will be decontaminated and remediated to the extent necessary to meet the license termination criteria. These include the decks and bulkheads which form the boundaries of radiologically controlled areas, and the contaminated liquid storage tanks that are integral to the ships double bottom hull structure.

At the end of the project, the final configuration of the ships former RCAs is expected to be as follows:

All RCAs (including low-level radioactive waste (LLRW) storerooms, the material handling area in Cargo Hold 4, and controlled areas in Cargo Hold 3) will be de-posted and removed from radiological controls as part of the License Termination process. The end-state configuration of these spaces will be unrestricted release from radiological controls.

The Containment Vessel and Secondary Shielding will be intact and retained in-situ. The Containment Vessel Cupola Head and Shield Ring will be reinstalled. Access (gratings, platforms and ladders) and lighting within the containment vessel will be restored.

The Reactor Compartment Hatch will be closed, and its rigging system will be removed and stowed in Cargo Hold 1. Once complete, the hatch will be inoperative without the use of an exterior crane.

The Decommissioning Heating Ventilation and Air Conditioning system described in the NS Savannah Updated Final Safety Analysis Report (UFSAR), Revision 13, dated July 20, 2023 (ML23208A038), will be decontaminated to the extent necessary and retained in operating condition.

Section 3.1.2, Completed Dismantlement Activities, of the LTP explains that major dismantlement activities were completed in the first half of CY 2023. Previously completed dismantlement activities have been summarized in each calendar year annual report since dismantlement was authorized by License Amendment 15 issued on April 23, 2018 (ML18081A134).

The LTP continues to explain that during D&D operations, the licensee concluded that there was potential preservation value to retaining the aft Reactor Coolant System (RCS) piping, and deferred shipping of this waste material until sampling and analysis was completed with a view towards decontamination of the pipe segments to levels below the DCGLs. The aft RCS piping segments were moved to a temporary radiologically managed area created in Cargo Hold 3 pending evaluation. Metallurgical samples were shipped in the third quarter of 2023. At the time

28 of LTP submittal, the licensee has not yet decided the fate of the pipe segments. If disposed, these segments will be among the last LLRW shipments. All dismantlement was essentially complete at the time of LTP submittal.

3.2 Remaining Site Dismantlement Activities Section 3.2, Remaining Activities, of the LTP states that decontamination and remediation of structures will be ongoing and will include those structures listed in Section 3.1.1 of the LTP.

These activities will continue, as required, during the FSS period.

Section 3.1.1, Dismantlement Scope and Planned Final Ship Configuration, of the LTP describes certain structures that are planned for retention subject to confirmation that they meet the license termination criteria. If these structures fail to meet the criteria of the approved LTP, they will be dismantled and shipped as LLRW. This would require a period of remobilization by the licensees decommissioning contractor; consequently, no changes to the current RCA boundaries for personnel entry and exit, monitoring stations, dosimetry issue and material handling arrangements will be made until the status of the retained structures and components is confirmed by NRC. If required, a revision to the LTP will be submitted by the licensee.

Section 3.1.3, Coordination of Activities and Unreviewed Safety Questions, of the LTP explains that there are no remaining decommissioning activities which require specific coordination with other Federal or State agencies. The licensee has coordinated prior completed activities with such agencies, including the transit of the RPV through the Howard Street Tunnel beneath Baltimore City. The licensee maintains contacts with Federal, State, and local Baltimore City agencies as a matter of routine. Periodic updates will be provided as the project progresses to license termination.

Section 3.3, Waste Projections, of the LTP discusses the volume of radioactive waste related to decommissioning. Table 3-1, Projected Remaining Waste Quantities as of September 30, 2023, in the LTP estimates a total of 2345.5 cubic feet of radioactive waste remaining. Table 3-2, Summary of Waste Disposed from Ship through September 30, 2023, in the LTP provides a summary of the waste shipments through August 31, 2023. The total volume of LLRW for disposal was originally estimated at 22,844 cubic feet. The licensee decided not to process waste onsite or to attempt to segregate waste streams. The solid waste was, and future solid waste will be, shipped to the licensed EnergySolutions radioactive waste disposal facility in Clive, Utah. Liquid waste and future liquid waste will be shipped to EnergySolutions Erwin, TN.

Section 3.4 Occupational Exposure, of the LTP discusses the estimated dose for completing the remaining activities. Table 3-3, Radiation Exposure - Project Total and Estimate to Complete Remaining Activities, estimates a total of 2.178 person-rem. The licensee has stated that no quantity of radioactive material will be released to unrestricted areas during the completion of the scheduled remaining tasks.

The LTP explains that decommissioning activities at NS Savannah will continue to be conducted in accordance with the requirements of 10 CFR 50.82(a)(6) and (a)(7). At the time of LTP submittal, the remaining activities do not involve any unreviewed safety questions or changes to the Technical Specifications. If an activity requires prior NRC approval under 10 CFR 50.59, or a change to the technical specifications or license, a LAR will be made to the NRC for review and approval before implementing the activity in question.

29 3.3 NRC Evaluation of Remaining Site Dismantlement Activities The NRC staff reviewed the information in the NS Savannah LTP, for the ship in accordance with Section 2.3, Identification of Remaining Site Dismantlement Activities, of NUREG-1700, Revision 2. As described therein, the purposes of the NRC staffs review are to ensure the LTP (1) discusses the remaining tasks associated with D&D, estimates the quantity of radioactive material to be shipped for disposal or processing, describes the proposed control mechanisms to ensure that areas are not re-contaminated, and contains occupational exposure estimates and radioactive waste characterization; (2) describes the remaining dismantlement activities in sufficient detail to identify any associated inspection or technical resources that will be needed; (3) is sufficiently detailed to provide data for use in planning further D&D activities, including decontamination techniques, projected schedules, costs, waste volumes, dose assessments, and health and safety considerations; and (4) lists the remaining activities that do not require any additional licensing action.

The LTP summarizes the remaining site D&D activities and techniques to be used and includes information regarding those areas and equipment that need further radiological remediation and an estimate of radiological conditions that the licensee may encounter. The licensee provided a description of the major remaining components of radiologically contaminated systems and specific equipment remediation considerations and a general schedule for completion of the D&D milestones. The LTP also describes the proposed control mechanisms to ensure remediated areas are not re-contaminated. The LTP also includes estimates of associated occupational radiation doses, projected volumes of radioactive waste, and a description of radioactive waste characterization. The licensee provides a description of how the remaining D&D activities are evaluated against the licensing basis for the plant and the requirements in the license for any unreviewed safety questions.

Based on this review, the NRC staff determined that the licensee has identified, in sufficient detail, the remaining dismantlement activities necessary to complete decommissioning of the facility per 10 CFR 50.82(a)(9)(ii)(B) and provided a basis for use in planning further decommissioning activities to allow for the NRC to identify any inspection or technical resources needed during the remaining dismantlement activities.

4 PLANS FOR RADIOLOGICAL SITE REMEDIATION In accordance with 10 CFR 50.82(a)(9)(ii)(C) and the guidance of RG 1.179 and NUREG-1700, Chapter 4, Remediation Plans, of the LTP details the remediation methods and techniques that the licensee will use to demonstrate that the facility D&D activities will be conducted in accordance with established Radiation Protection, Safety, and Waste Management programs.

The licensee stated that the remaining residual radioactivity will satisfy the ALARA criterion in 10 CFR 20.1402.

Chapter 4, Remediation Plans, of the LTP describes plans for site remediation and discusses how facility and site areas will be remediated to meet the license termination criteria for unrestricted release, as specified in Subpart E of 10 CFR Part 20, including the proposed residual radioactivity levels defined as the DCGLs. The Remediation Plan is a list of techniques, methods and technologies that will be used to meet the two radiological criteria for unrestricted use specified in 10 CFR 20.1402: 1) The Total Effective Dose Equivalent (TEDE) from residual radioactivity that is distinguishable from background radiation must not be greater than 25 mrem/yr to the average member of the critical group; and 2) Residual radioactivity levels must

30 be ALARA. The effectiveness of these techniques will be confirmed by implementing the FSS plan described in Chapter 5 of the LTP.

NUREG-1700 recommends that the following be discussed in the LTP to evaluate whether the DCGLs have been met:

Summarize the techniques that will be used to remediate building structures and components.

Summarize the equipment that will be decontaminated and how the decontamination will be accomplished.

Summarize the radiation protection methods and control procedures that will be employed including a summary of the procedures already authorized under the existing license and any changes in the radiological controls to be implemented to control radiological contamination associated with the remaining decommissioning and remediation activities.

Commit to conduct decommissioning activities in accordance with approved written procedures.

Include a detailed description of the techniques that will be employed to remove or remediate surface and subsurface soils, groundwater, and surface water and sediments.

Describe plans, if any, for onsite disposal of decommissioning waste.

The LTP includes a schedule that demonstrates how and in what time frames the licensee will complete the interrelated decommissioning activities. The regulation in 10 CFR 50.82(a)(3) requires completion of decommissioning within 60 years. If the completion of decommissioning is delayed for more than 60 years, the LTP must include a justification for the delay in accordance with 10 CFR 50.82(a)(3).

The LTP provides that decontamination and dismantlement activities are conducted in accordance with NS Savannah administrative programs and procedures. These programs and procedures are frequently assessed for technical content and compliance. Revisions have been, and will continue to be made, to these programs and procedures to accommodate the changing work environment inherent to reactor decommissioning. The revisions will continue to be documented, processed, and approved in accordance with the existing NS Savannah license, Technical Specifications, and Administrative Procedures. Consistent with RG 1.179, details regarding changes to the RPP to address remediation and decommissioning activities are not provided in this LTP. Changes to the RPP will be provided in either 1) Annual Reports as required by the Technical Specifications or 2) in periodic updates to the UFSAR or LTP.

As described in Sections 2.1.4 and 5.2 of the LTP, the NS Savannah site contains no surface or subsurface soils, groundwater, or surface water features, and consequently, there is no detailed description in the LTP of any techniques that would normally be employed to remove or remediate them.

4.1 NS Savannah Remediation Plans The licensee plans to remediate the site, including structures and systems that remain on the site, to the criteria of 0.15 mSv/yr (15 mrem/yr) for all pathways. However, the NRC staff has evaluated the licensees proposed LTP, including its associated dose assessments, against the 25 mrem/yr dose limit in 10 CFR 20.1402. The 15 mrem/yr adopted dose standard has been evaluated as an additional safety margin and provides additional assurance that the licensee will be successful in meeting the established DCGLs for license termination.

31 Section 4.2, Remediation Actions, of the LTP provides that remediation actions are performed throughout the decommissioning process and are based on the results of radiological surveys.

For example, a survey may be performed in an area where prior to a large component removal, no survey was possible and the collected survey data indicates that remediation is required.

When decommissioning activities in a survey unit are complete, characterization surveys are conducted when additional information is needed for the proper planning of the survey units FSS. A turnover survey will be performed in preparation for FSS activities when a review of previous survey data for the survey unit cannot confirm a) there is sufficient data to perform FSS design and b) an FSS in the survey unit is likely to meet release criteria. If the characterization survey or turnover survey determines that additional remediation is necessary, then the completed remediation activities will be followed by a Remedial Action Support Survey to verify that a survey unit is likely to meet release criteria following the performance of an FSS.

Remedial action support survey design is described in Sections 5.3 and 5.4 of the LTP.

The LTP further provides that the remediation techniques, methods, and technologies planned to be used by the NS Savannah are standard to the commercial nuclear industry. They represent the current best practice methods and use a consistent approach intended to facilitate the most cost-effective balance between hazardous waste removal and decontamination cleaning. All remediation actions described may not necessarily be required but are listed as possible actions that may be taken during the decommissioning of the NS Savannah. The appropriate remediation technique(s), method(s) and/or technologies that will be employed are dependent on the physical composition and configuration of the contaminated media requiring remediation. At the NS Savannah, the principal media that will be subjected to remediation are structural surfaces. There has been minimal contamination identified from characterization survey results and historical survey data.

The licensee describes the following techniques, methods, and technologies that may be used during remediation:

Pressure Washing: removes superficial materials from the suspect surface; Needle Guns: a method of scrabbling used for removal and chipping of media is usually reserved for areas not accessible to normal scabbling operations; High Pressure Water Blasting: may be used to remediate the pipe interior surfaces; Laser Ablation: used to dislodge surface contamination such as oxides and coatings from the surface of a substrate. This system has been used to decontaminate some of the inside surfaces of the Pressurizer; Chemical Strippers: may be used for the removal of certain contaminants in small areas; Grinding: used to remove material from surfaces, cracks, or corners.;

Sponge and Abrasive Blasting: used for the removal of surface films and paints; Chemical decontamination and ultrasonic cleaning for select sections of the RCS piping.

Pipes, surfaces and drain lines can be cleaned and hot spots removed using these techniques and technologies.

Section 4.3, Remediation Activities Impact on the Radiation Protection Program, of the LTP stated that the current RPP is adequate to safely control the radiological aspects during decommissioning and does not present new challenges above those encountered during normal plant operations. The licensee states the program is protective of occupational personnel expected to encounter radiological hazards from decommissioning, ensures the protection of the public from radiological hazards, and makes sure occupational, effluent, and environmental

32 dose from radiological materials remain ALARA. Furthermore, it also stated that during decommissioning, contamination reduction techniques and engineering controls are used to mitigate the spread of contamination and reduce personnel exposure to radiation and contamination.

Section 5.4.4, Area Preparation: Isolation and Control, of the LTP describes how the licensee plans to mitigate the spread of contamination once an area is remediated. Before FSS activities can begin in an area, a transition must occur where planned decommissioning activities are completed, and the area is subsequently assessed to scope the required isolation and control measures. A walkdown will occur to establish if the area is ready for final survey activities and identify any work practice issues that must be addressed in survey planning and design.

Determination of readiness for FSS will be based on a characterization survey, turnover survey and/or a Remedial Action Support Survey (RASS) indicating that the residual radioactive material is likely to comply with the FSS criteria.

The licensee describes the following criteria must be met for an area to be deemed ready for isolation and control:

Known contaminated decommissioning activities in the area are complete and any additional decommissioning activities identified shall pose a very low risk to add contamination to an area, including removal, as necessary, of items (e.g., equipment mounts, bulkhead hangers, and exposed studs) that could interfere with final survey activities; All planned decommissioning activities in areas either adjacent to the survey unit to be isolated or that could otherwise affect it are controlled using isolation and control techniques, are complete or are deemed not to have any reasonable potential to spread radioactive material to the survey unit; All tools and equipment not needed for final survey activities are removed; Any equipment to be used for final survey activities is evaluated to ensure it does not pose the potential for introducing radioactive material into the survey unit; and, Where practical, transit paths to or through the survey unit, except those required to support final survey activities, are eliminated or re-routed.

The LTP section concludes in stating that once the area meets the isolation and control criteria, isolation and control will be achieved through a combination of personnel training, physical barriers, postings, and site notices as appropriate, to prevent unauthorized access to an isolated survey unit. Isolation and control measures will be implemented through approved plant procedures. An administrative process will be used to evaluate, approve (or deny), and document all activities conducted in these areas during and following FSSs.

In its letter dated September 16, 2024, the NRC staff asked the licensee to provide criteria for surveillance and surveys that must be met and what corrective actions will be taken if criteria are not met. In its RAI response dated October 16, 2024, the licensee stated in response RAI 5-8, that procedure STS-005-032, Survey Unit Turnover and Control, was revised to require the performance of periodic surveillances on completed survey units.

4.2 NRC Evaluation of the Radiological Site Remediation Plan The NRC staff evaluated the information in the LTP in accordance with Section 2.4, Remediation Plans, of NUREG-1700, Revision 2. As described therein, the purposes of the

33 NRC staffs review are to ensure the LTP: (1) addresses any changes in the radiological controls to be implemented to control radiological contamination associated with the remaining decommissioning and remediation activities; (2) discusses in detail how facility and site areas will be remediated to meet the proposed residual radioactivity levels (DCGLs) for license termination; and (3) includes a schedule that demonstrates how and in what time frame the licensee will complete the interrelated decommissioning activities.

The LTP discusses in detail how the licensee intends to remediate the ship to meet the proposed residual radioactivity levels (DCGLs) for license termination, including a summary of the removal and remediation tasks planned for structures at the site, as well as the techniques associated with these tasks. The LTP also includes a summary of the radiation protection methods and control procedures that will be employed during the remaining decommissioning activities. The LTP provides the details of the licensees ALARA analyses to ensure compliance with the criterion specified in 10 CFR 20.1402. Table 3-4, General Project Milestones of the LTP contains a schedule that demonstrates how the licensee intends to complete the interrelated decommissioning activities.

The NRC staff evaluated the licensees ALARA analyses in Section 5.3.2 of this SER. The 25 mrem/yr dose criteria of 10 CFR 20.1402 will apply even though the licensee will have a self-imposed administrative limit at a lower dose level. Compliance with the NRCs regulations, including ALARA, will be achieved when demonstrating compliance with the 25 mrem/yr dose criteria and implementing good ALARA practices at the site as outlined in NUREG 1757. The licensees commitment to this criteria and good ALARA practices is set forth in Section 5.3.2 of this SER. The NRC staff finds that the proposed commitments will be adequate to ensure compliance with 10 CFR 20.1402 with respect to meeting ALARA.

Therefore, the NRC staff finds that the NS Savannah LTP meets the acceptance criteria as delineated in Section 2.4 of NUREG-1700, and Section 6.3.6.2, Performance-Based Compliance, in NUREG-1757 which is one way to meet the regulatory requirements. Based on this review, the NRC staff finds the licensee has provided a sufficiently detailed discussion of its radiological site remediation plans for the remaining decommissioning activities, as required by 10 CFR 50.82(a)(9)(ii)(C).

5 FINAL STATUS SURVEY PLAN The FSS is the radiation survey performed after an area has been fully characterized and remediated and the licensee believes that the area is ready to be released. The purpose of the FSS is to demonstrate that the site, or portion thereof, under consideration meets the radiological criteria for license termination in Subpart E of 10 CFR Part 20.

In accordance with the requirements of 10 CFR 50.82 (a)(9)(ii)(D) and the guidance of RG 1.179, and NUREG 1700, Chapter 5, Final Status Survey Plan, Chapter 5 of the LTP provides a description of the methods to be used in planning, designing, conducting, and evaluating the FSS at the NS Savannah. The FSS plan describes the final survey processes that will be used to demonstrate that the NS Savannah complies with the radiological criteria for unrestricted use specified in 10 CFR 20.1402. Additional regulations applicable to FSS are also found in Subpart F, Surveys and Monitoring, of 10 CFR Part 20 at 10 CFR 20.1501(a) and (b).

According to the guidance contained in RG 1.179, and outlined by the LTP in Section 5.1, Introduction, a licensee should include the following items, which are not meant to be all-inclusive, in the final radiation survey plan:

34 Describe the methods proposed for surveying all equipment, systems, structures, and soils, as well as a method for ensuring that sufficient data are included for a meaningful statistical survey.

Describe the methods the licensee will use to establish background radiation levels.

Include a discussion of variances in background radiation that can be expected (e.g.,

between structures constructed of different materials).

Describe the QA program to support both field survey work and laboratory analysis.

Address the QA organization; training and qualification requirements; survey instructions and procedures, including water, air, and soil sampling procedures; document control; control of purchased items; inspections; control of survey equipment; handling, storage, and response checks; shipping of survey equipment and laboratory samples; disposition of nonconformance items; corrective action; QA records; and survey audits, including methods to be used for reviewing, analyzing, and auditing data.

Describe the verification surveys and evaluations used to support the delineation of radiologically affected (contaminated) areas and unaffected (uncontaminated) areas.

Identify the major radiological contaminants.

Discuss methods used for addressing hard-to-detect radionuclides.

Describe access control procedures to avoid recontamination of clean areas.

Identify survey units having the same area classification.

Describe scanning performed to locate small areas of elevated concentrations of residual radioactivity.

Discuss levels established for investigating significantly elevated concentrations of residual radioactivity.

Describe the reference coordinate system established for the site areas.

Chapter 5 of the NS Savannah LTP notes that the FSS plans were based on multiple guidance documents including: RG 1.179; NUREG-1700, Revision 2; NUREG-1757, Volume 2, Revision 2; NUREG-1575, Revision 1; NUREG-1505, A Non-parametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys, Revision 1, (ML061870462); and NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, Revision 1, (ML003676046).

Chapter 5 of the NS Savannah LTP provides a description of the methods used in planning, designing, conducting, and evaluating the FSS for NS Savannah. The FSS plan describes the final survey processes that will be used to demonstrate that the NS Savannah complies with the radiological criteria for unrestricted use specified in 10 CFR 20.1402. Additional regulations applicable to FSS are also found in Subpart F of 10 CFR Part 20 and 10 CFR 20.1501 (a) and (b).

The licensee developed a FSS design using the Data Quality Objective (DQO) process outlined in Appendix D of MARSSIM. The seven steps for the DQO process are discussed in Section 5.4.1, Data Quality Objectives, and its subsections of the LTP. The general approach in MARSSIM for FSS involves a statistically determined number of measurements or samples being taken within a survey unit so that the non-parametric statistical tests used for data assessment can be applied with adequate confidence. For the NS Savannah FSS, the licensee chose the Sign Test as the non-parametric statistical test because the background is expected to constitute a small fraction of the DCGLW based on results of characterization surveys.

35 Section 5.2, Scope, of the LTP states that the scope of the NS Savannah FSS plan encompasses the radiological assessment of all affected structures and systems in the ship for the purpose of quantifying the concentration of any residual activity that remains following all decontamination activities. Concentration limits have been established to represent the maximum dose rate criterion for unrestricted release specified in 10 CFR 20.1402 and are presented in Chapter 6 of the LTP. The NS Savannah possesses unique features that both simplify, and perhaps complicate, the FSS process as compared to typical land-based sites.

Section 5.3, Summary of the Final Status Survey Process, of the LTP discusses an overview of survey unit design. The licensee plans to utilize the DQO and data quality assessment processes for systematic planning and to address problems that require a decision to be made.

Survey planning includes review of the HSA and other pertinent characterization information to establish the ROCs and survey unit classification. For the NS Savannah, FSS data will only be evaluated with the Sign Test because the radionuclide(s) of interest either does not exist in background or is not present in a concentration that is a relevant fraction of the DCGLW. QA and quality control (QC) measures will be employed throughout the FSS process to ensure that all decisions are made based on data of acceptable quality.

The survey design selects the appropriate survey instruments and techniques to provide adequate coverage of the unit through a combination of scans, fixed measurements, and sampling. This process ensures that data of sufficient quantity and quality are obtained to make decisions regarding the suitability of the survey design assumptions and whether the unit meets the release criterion. A reference coordinate system may be used for documenting locations where measurements were made and to allow replication of survey efforts if necessary.

Approved procedures will direct this process to ensure consistent implementation and adherence to applicable requirements. Survey implementation will be the process of performing the survey plan for a given survey unit. This will consist of scan measurements and fixed measurements. Data will be stored and controlled. Survey results will be converted to appropriate units (i.e., dpm/100cm2) and compared to investigation levels to determine appropriate follow-up action.

The LTP section concludes in stating that documentation of the FSS survey will transpire in two types of reports. An FSS Survey Unit Release Record will be prepared to provide a complete record of the "as left" radiological status of an individual survey unit, relative to the specified release criteria. An FSS Final Report, which is a written report submitted to the NRC, will be prepared to provide a summary of the survey results and the overall conclusions from multiple survey units. It will include the associated Release Records. These reports will demonstrate that the NS Savannah meets the radiological criteria for unrestricted use.

5.1 Determination of FSS Data Requirements 5.1.1 Survey Unit Classifications and Area Limitations Section 5.4.2, Survey Units, of the LTP discusses the survey unit classification. Table 5-1, Survey Unit Surface Area Limits, in the LTP presents the limits for survey unit areas that are classified as impacted to facilitate survey design. Each survey unit has been assigned an initial classification based on the site characterization process and the HSA as described in Chapter 2 of the LTP. The licensee does acknowledge in the LTP that situations may arise where it is necessary to create new survey units by subdividing areas within an existing unit or changing the classification of a survey unit if the contamination detected in a survey unit exceeds the general category for the survey unit classification.

36 5.1.2 Reference Coordinate System Section 5.4.3, Reference Coordination Systems, of the LTP discusses the coordinate system that will be utilized for mapping and planning the surveys. On a ship, the location of any given point is defined by identifying its distance from one of three axes or reference points. Elevation is typically defined by the deck (floor) on which the point exists, or by its height above baseline.

The baseline is a horizontal line that follows the molded line of the keel and represents the bottom of the ship. Because some exterior and interior decks have an upward rise known as sheer, the height of a particular deck above baseline may vary along the ships length.

The point will be defined by its distance from the Forward Perpendicular. The Forward Perpendicular is an arbitrary fixed vertical line that runs through the intersection of the ships bow with its design waterline. The ships transverse frames are numbered running aft, with the Forward Perpendicular having no frame designation, but essentially being equal to Frame 0.

Transverse bulkheads (walls) are typically aligned on frames.

Finally, the LTP provides that the point is defined by its distance from the ships Centerline. The Centerline is a vertical plane running the length of the ship. Distances are measured to port (left) or starboard (right). Ship compartments (rooms or spaces) may be numbered using an alpha-numeric designation to represent its location relative to the three axes (Deck, Frame, and Center). Compartments that are to port of center use an even number designation.

Compartments that are to starboard of center use an odd number designation. Compartments which extend to both sides of the Centerline are said to be on Center and use the numeral 0 in their designation. However, on the NS Savannah, many compartments employ a simple noun-name designation.

The LTP section concludes in stating that the reference coordinate system for NS Savannah survey units will include a benchmark, or origin point. This benchmark will be defined using the ships reference system and will typically be located at the lowest, forwardmost point of the survey unit that is closest to the Centerline. The coordinate system used for surveys will typically take the form of a triangular grid of intersecting, perpendicular lines; but other patterns (e.g., rectangular and polar) may be used as convenient, depending on the circumstances or shape of the survey unit. The system will serve as a convenience for documenting survey efforts and other information pertaining to a given survey unit. The coordinate system also provides a means to specify general locations for measurements performed for QC or verification purposes. The benchmark and survey pattern will be provided for each survey unit in the FSS packages. Physical gridding of a survey unit will only be done in cases where it is beneficial and cost-effective to do so. When physical gridding is used, benchmark locations will be designated by marking a spot by a suitable technique.

5.1.3 Determining Measurement Locations Section 5.5.1.4, Determining Measurement Locations, of the LTP states that, for Class 1 and Class 2 survey units, fixed measurements will be performed over a systematic measurement pattern consisting of a grid having either a triangular or a square pitch. The pitch (grid spacing) will be determined based on the number of measurements required and whether the desired grid is triangular or square. Given that a triangular grid in general is more efficient than a square grid for detecting small areas of elevated activity, triangular grids should be employed for FSSs involving fixed measurements in Class 1 and Class 2 survey units when practical.

37 Systematic grids will not be used for surveys involving fixed measurements for Class 3 units.

Instead, fixed measurement locations will be selected at random throughout the survey unit area by generating pairs of random numbers between zero and one. One pair of random numbers will be generated for each fixed measurement to be made. The random number pairs, representing (x,y) coordinates, will be multiplied by the maximum length and width dimensions of the survey unit to yield the location for each fixed measurement. For odd-shaped survey units, a rectangular area encompassing the survey unit will be used to establish the maximum length and width. A new pair of random numbers will be generated if any of them give locations that are not actually within the survey unit boundaries. New pairs of numbers will also be generated in cases where a measurement cannot be made at a specific location because of an obstruction, inaccessibility, etc. The spacing to be used in setting up the systematic grid used to establish fixed measurement locations for Class 1 and Class 2 areas will be computed using equation 5-4 for a triangular grid, or Equation 5-5, for a square grid.

Once the grid spacing is established, a random starting point will be established for the survey pattern using the same method as described above for selecting random locations for Class 3 units. Starting from this randomly selected location, a row of points will then be established parallel to one of the survey unit axes at intervals of L. Additional rows will then be added parallel to the first row. For a triangular grid, additional rows will be added at a spacing of 0.866L from the first row, with points on alternate rows spaced mid-way between the points from the previous row. For a square grid, points and rows will be spaced at intervals of L. Section 5.5.2.5 of MARSSIM describes the process to be used for selecting fixed measurement locations and provides examples of how to establish both a systematic grid and random measurement locations.

The LTP concludes in stating that measurement locations may also be determined with VSP.

Pacific Northwest National Laboratory created VSP. VSP is a software tool that supports the development of a defensible sampling plan based on statistical sampling theory and the statistical analysis of sample results to support confident decision-making. VSP couples the site, building, room and sample location visualization capabilities with optimal sampling design and statistical analysis strategies.

5.1.4 Estimating the Number of Measurements Required Section 5.5.1, Selecting the Number of Fixed Measurements and Locations, of the LTP states that the MARSSIM methodology for evaluating whether a survey unit meets its applicable release criterion using fixed measurements plus scans is based on using non-parametric statistical tests for data assessment. Selection of the required minimum number of data points depends on which statistical test is going to be used to evaluate the data, and thus depends on what type of measurements are to be made (gross measurement, net measurement or radionuclide specific) and if the radionuclide(s) of interest appear(s) in background. On the NS Savannah, FSS data will only be evaluated with the Sign Test because the radionuclide(s) of interest either does not exist in background or is not present in a concentration that is a relevant fraction of the DCGLW.

Section 5.5.1.1, Establishing Acceptable Decision Error Rates, of the LTP states that the decision error rates for FSSs designed for the NS Savannah site will be set as follows:

The value (Type 1 error) will always be set to 0.05 unless prior NRC approval is granted for using a less restrictive value; and,

38 The value (Type 2 error) is nominally set to 0.05 but may be changed without NRC approval if it is found that more fixed measurements than necessary are being made to demonstrate compliance with the release criterion.

Section 5.5.1.2, Determination of Relative Shift, of the LTP states that the relative shift is a parameter that quantifies the concentrations to be measured in a survey unit relative to the variability in these measurements. The relative shift is a function of the DCGLW, a parameter called the lower bound of the gray region (LBGR) and the expected standard deviation of the measurements to be made in the survey unit (s). The s values will be selected by:

Using existing characterization or remediation support survey data; or, Making preliminary measurements.

Given that s values should reflect a combination of the spatial variability in the concentration and the precision in the method of measurement, these values will be selected based on existing survey data only when the existing measurements were made using techniques equivalent to those to be used during the FSS.

According to the LTP, the LBGR represents the concentration to which the survey unit must be cleaned (decontaminated) to have an acceptable probability of passing the statistical test. The difference between the DCGLW and the LBGR, known as the shift, can be thought of as a measure of the resolution of the measurements that will be made in a survey unit. If the LBGR is near the DCGLW, the shift will be small, and thus a strong potential for Type 1 errors will exist.

Likewise, if the shift is large, the probability of Type 2 errors increases. The shift is denoted as The relative shift (/ s) is computed as the quotient of the shift and the appropriate standard deviation values. If no reference area data are needed to evaluate the survey results, the expected standard deviation of the measurements (s) is used. When preliminary data are not obtained, it may be reasonable to assume a coefficient of variation on the order of 30%, based on experience.

To compute the relative shift, the appropriate sigma value and an initial LBGR are selected. Per MARSSIM, the initial value for the LBGR will be set to one-half of the DCGLW. If the resulting relative shift is not between 1.0 and 3.0, the LBGR is adjusted until it is. If the relative shift is too low, the LBGR is decreased; if the relative shift is too high, the LBGR is increased.

Section 5.5.1.3, Selecting the Required Number of Measurements for the Sign Test, of the LTP states the minimum number of fixed measurements required for the Sign Test is computed by Equation 5-3 in the LTP, copied and pasted below. This equation is a duplicate of equation 5-2 in MARSSIM for determining the number of data points for the Sign Test.

39 In lieu of calculating the value of N (the minimum number of measurements required) by Equation 5-3, the value of N will be obtained from Table 5.5, Values of N for Use with the Sign Test in MARSSIM. On NS Savannah, FSS data will only be evaluated with the Sign Test because the radionuclide(s) of interest either does not exist in background or is not present in a concentration that is a relevant fraction of the DCGLW.

5.1.5 Scanning Section 5.5, Final Status Survey Design Elements, of the LTP discusses the general survey designs. The coverage requirements that will be applied for scans performed in support of FSSs for the NS Savannah are:

For Class 1 survey units, 100% of the accessible surface will be scanned; For Class 2 survey units, between 10% and 100% of the surface will be scanned in a combination of systematic and judgmental measurements for external surface area units and for deck and lower bulkheads of structures; and 10% to 50% of the surface will be covered for upper bulkheads and overheads; Scanning will be done on a judgmental basis for Class 3 survey units.

Section 5.5.3.1, Investigation Levels, of the LTP presents the investigation levels for fixed measurements and scanning surveys. Investigation levels are determined during survey planning to identify areas of possible elevated activity. Scanning surveys are performed to locate contamination anomalies indicating residual gross activity that may require further investigation or action. Table 5-2, Investigation Levels, in the LTP presents the investigation levels proposed by the licensee.

In its letter dated September 16, 2024, the NRC staff asked the licensee to clarify its June 27, 2024, response to RAI 5-3, which requested confirmation of plans to remediate materials with measurements above the DCGLW until residual radioactivity is less than the DCGLW to address demonstrating compliance with 10 CFR 20.1402. In its RAI response dated October 16, 2024, the licensee revised Table 5-2 in the response to RAI 5-3 to clarify the criteria for the different Class survey units as copied and pasted below. In that response, the licensee states that all external surfaces above the DCGLW will be investigated and remediated or removed to less than the DCGLW. Any internal surface in a system above the DCGLW will be remediated, or the component will be removed and properly disposed. When an investigation level is exceeded, the first step is to confirm that the initial measurement result exceeds the

40 particular investigation level. Depending on the results of the investigation actions, the survey unit may subsequently require re-classification, remediation, and/or re-survey. Investigation levels are established for each class of survey unit.

Section 5.7.1, Survey Methods, of the LTP discusses scanning and the requirements for doing scanning surveys. Table 5-3, Traditional Scanning Coverage Requirements, Survey Unit Classification, Required Scanning Coverage Fraction, in the LTP presents the area coverage requirements when scanning used with fixed measurements as copied and pasted below.

5.1.6 Additional Fixed Measurement Commitments General fixed measurement commitments are generally discussed in the FSS design section of this SER with respect to the number of measurements anticipated in each survey unit and the limits for survey unit size. This section will further discuss judgmental measurements, surveys of systems and components, investigation of elevated areas and similar topics pertinent to the surveys.

Section 5.4.5.3, Elevated Measurement Comparison and Area Factors, of the LTP discusses the criteria for identifying and assessing areas of surveys units which exceed the DCGLW. As discussed in 5.5.3.1, Investigation Levels, of the LTP for Class 1 survey units, measurements above the DCGLW are not necessarily unexpected. However, such a result may still indicate a need for further investigation if it is significantly different than the other measurements made within the same survey unit. Measurements in Class 1 survey units that exceed the DCGLW and differ from the mean of the remaining measurements by more than three (> 3) standard

41 deviations will therefore be investigated. Measurements in Class 1 units that exceed the DCGLW, but do not differ from the mean by less than or equal to three ( 3) standard deviations may still be investigated based on professional judgment, as may any measurements that differ significantly from the rest of the measurements made within a given survey unit. Therefore, no area factors have been calculated and the DCGLEMC will not be used.

Section 5.5.2, Judgmental Assessments, of the LTP discusses judgmental measurements, or biased measurements where the licensee states it plans to perform biased measurements beyond those required to investigate exceedances of the investigation levels discussed in Table 5-2. In Class 1 survey units, judgmental measurements will be based on the scanning measurements and the suggested investigation levels. However, in Class 2 and 3 survey units, biased measurements will be performed in locations based on site knowledge and professional judgment. The basis for the judgmental assessments will be documented in the survey package for each survey unit.

Section 5.5.3.2, Investigations, of the LTP states that any exceedance of an applicable investigation level will be reassessed to confirm the measurement. If confirmed, additional measurements will be made to determine the extent of the area of elevated activity and to provide reasonable assurance that other areas of elevated activity do not exist. If residual activity in excess of the applicable investigation level is confirmed, the documentation will also address the potential source(s) of the activity and the impact this has on the original classification assigned to the survey unit. A decision will then be made regarding re-classification of the unit in whole or in part.

Section 5.5.3.3, Remediation, of the LTP states that if, during the performance of an FSS, any areas of residual activity are found to be in excess of the DCGLW and an outlier, those areas will be remediated with the goal to reduce the activity to less than or equal to the DCGLW. In its letter dated May 30, 2024, the NRC staff asked in RAI 5-3 that the licensee to confirm plans to remediate materials with measurements above the DCGLW until residual radioactivity is less than the DCGLW or provide instructions on how any elevated measurements will be addressed in demonstrating compliance with 10 CFR 20.1402. In its RAI response dated June 27, 2024, the licensee states that all external surfaces above the DCGLW have been or will be remediated to less than the DCGLW. Any internal surface in a system above the DCGLW will be remediated or removed and properly disposed. The investigation level is set at 75% of the DCGLW which provides confidence that measurement locations greater than the DCGLW will either be remediated or removed.

The LTP further states that in Class 2 or Class 3 areas, neither measurements above the DCGLW nor areas of elevated activity are expected. Thus, any fixed measurements or sampling results that exceed the DCGLW in these areas will be investigated. In the case of Class 3 areas, where any residual radioactivity would be unexpected, fixed measurement or sample results that are greater than 0.25 x DCGLW will be investigated.

Section 5.5.3.5, Re-survey, of the LTP states that if a survey unit is re-classified (in whole or in part), or if remediation is performed within a survey unit, then the areas affected are subject to re-survey. Any re-surveys will be designed and performed based on the appropriate classification of the survey unit. That is, if a survey unit is re-classified or a new survey unit is created, the survey design will be based on the new classification. If a survey unit is sub-divided, the survey design for the remaining area of the original survey unit may or may not be affected depending on the remaining surface area of the unit and its classification. If the original survey unit was Class 3, then the only impact on the survey design (in the case of fixed

42 measurements or sampling) is to perform additional measurements at randomly selected locations until the required total number of measurements is met. If the original survey unit was Class 2, the spacing of the measurement locations may need to be adjusted depending on the remaining surface area of the survey unit relative to its original area. If there is a large change in the surface area, then a new survey design will be necessary to accommodate the required number of measurements in the smaller area. If there is not a large change in area, then the impact on the grid spacing is minimal (with respect to areal coverage), and additional measurement locations need only be selected at random to obtain the required number of measurements.

5.1.7 Surveys for Non-Structural Systems and Components Section 5.6, Survey Protocol/or Non-structural Systems and Components, of the LTP provides that the guidance provided in MARSSIM for conducting FSSs does not include guidance for conducting FSSs for non-structural system or components. Examples of non-structural systems and components include pumps, motors, heat exchangers, and piping between components.

Surface activity assessments for non-structural systems and components may be made by making measurements at traps, tanks, open piping and other appropriate access points where activity levels should be representative of those on the interior surfaces. Assessments may also be made via in-situ gamma spectroscopy or pipe crawlers, provided adequate instrument efficiencies and detection limits can be achieved. Detection limits for surface activity assessments should be at least 50% of the release limits. If necessary, scaling factors may be applied to establish gross activity levels via radionuclide specific measurements or other assessments, as appropriate.

Section 5.9, Notes on Structure and System Surveys, of the LTP provides descriptions of plans and for select structures and systems. This includes: exterior hull surveys, neutron shield tank/fuel transfer tank surveys, steam generators, pressurizer, and the double bottom tanks. In all cases, the surveys are planned to be MARSSIM based and primarily incorporates measurements of the most likely contaminated portions of the structures (e.g., interior of tanks).

5.1.8 NRC Evaluation of FSS Data Requirements The NRC staff evaluated the information in Chapter 5 of the NS Savannah LTP against the acceptance criteria in Section 2.5, Final Radiation Survey Plan, of NUREG-1700. As described therein, the purpose of the NRC staffs review is to ensure the LTP includes (1) the Information To Be Submitted, as described in Section 4.4, Final Status Survey Design, of NUREG-1757, Volume 2, Revision 2; (2) the following information: identification of the major radiological contaminants; methods used for addressing HTD radionuclides; access control procedures to control recontamination of clean areas; description of the QA program, and; methods for surveying embedded and buried piping; and (3) a final survey plan that meets the evaluation criteria defined in Section 4.4.1.2 of NUREG-1757, Volume 2, Revision 2.

The NRC staff noted during the FSS that the licensee did not appear to take into consideration the surface efficiency when considering the determination of MDC. While this did not significantly impact the suitability of the instruments for use during the FSS, the NRC staff did communicate to the licensee that measurements should also be incorporating the surface efficiency into their determinations for the FSS.

The NRC staff finds that the FSS design plan is generally consistent with the guidance in NUREG-1700, Revision 2; NUREG-1757, Volume 2, Revision 2, and NUREG 1575. The only

43 exception appears to be that the licensee plans to only consider an elevation to occur if it both exceeds the DCGLW and is also an outlier. If an elevated area does exist, the licensee plans to remediate with the goal of decreasing the residual radioactivity to levels less than the DCGLW.

For this reason, the licensee also does not consider it necessary to establish area factors for establishing the DCGLEMC. The licensees DCGLWs are being established to demonstrate it is at a 15 mrem/yr level as opposed to the NRCs license termination criteria of 25 mrem/yr. Also, the residual radioactivity concentrations typically present after conclusion of remediation efforts are much less than the DCGLW so the NRC staff finds it extremely likely any exceedance of the DCGLW would also be an outlier necessitating the licensee to evaluate and further remediate.

As such the licensees modification of the investigation level and further remediation requirements are acceptable to the NRC staff, in this case.

5.2 Radionuclides of Concern Section 6.4, Radionuclides for Evaluation, of the LTP discusses the licensees anticipated ROCs and fractional makeup to be encountered during decommissioning. The licensee states that 12 composite smear samples taken during 2019 characterization of the Containment Vessel and Reactor Compartment were sent to GEL in South Carolina for gamma spectrometry.

Five of those composite smear samples were also analyzed for HTD radionuclides. In addition, a sludge sample was also sent for HTD analysis. In its letter dated September 16, 2024, the NRC staff asked the licensee a second round of RAIs to clarify the LTPs description as to how the ROCs were determined. In its RAI response dated October 16, 2024, the licensee included to the second round of RAI responses which clarified how the ROCs were determined.

5.2.1 Site ROC Determination The NRC staff determines, upon consideration of the LTP, that the NS Savannah structures and systems were potentially impacted as a result of reactor operation. Components or structures exposed to neutron fluence during operation could have become radiologically contaminated due to neutron activation of the elemental makeup within the volume of the structural system matrix and sediment or sludge buildups within reactor systems. Volumetric contamination should be limited to materials within the containment vessel, and the intent of the Neutron Shield Tank was to minimize activation beyond the reactor vessel. Impact to other surfaces could have occurred due to spills or leaks of coolant that had been contaminated from any fuel failures, such as the suspected failure noted during a fuel shuffle outage in 1968 or neutron activation of corrosion products within the coolant.

Table 6-2, Initial Suite of Radionuclides, in Section 6.4 of the LTP provides an initial list of potential radionuclides the licensee developed. The list was developed based on characterization campaigns and the collection of smear, paint, metal, concrete cores, residues, and water samples. The results of laboratory analysis of six samples sent for offsite analysis, including hard-to-detect radionuclides, are presented in Table 6-3, Offsite Laboratory Results of Sample Analyses, in the LTP. The anticipated primary ROCs consisted of the activation products tritium (H-3), carbon-14 (C-14), iron-55 (Fe-55), cobalt60 (Co-60), nickel-63 (Ni-63),

and silver-108m (Ag-108m) and fission products: strontium90 (Sr-90), technetium-99 (Tc-99),

and cesium-137 (Cs-137). Alpha emitters were not identified apart from americium-241 (Am-241) and plutonium-239/240 (Pu239/240) detection in a sludge sample from a makeup storage tank sludge. The licensee then determined each radionuclides fraction of the total activity in each sample in Table 6-4, Radionuclide Fractions of the Sample Analysis. For radionuclide results that were less than the MDA, the fraction of the total activity for that

44 radionuclide was assigned a value of (0). The fractions from Table 6-4 were summed and then normalized to obtain the normalized sum of fractions radionuclide fractions, which are presented in Table 6-5, Radionuclide Sum of Fractions and Normalized Sum of Fractions, in the LTP.

Section 6.13, Radionuclides of Concern and Insignificant Dose Contributors, of the LTP describes that after evaluating the activity and dose fractions for the various scenarios the licensee proposed for the NS Savannah, the licensee determined only two ROCs (Cs-137 and Co-60) are significant with the remainder being considered insignificant contributors. However, this is complicated somewhat in that Section 5.4.5.2, Surrogate Ratio DCGLs, of the LTP notes that no single radionuclide can be screened out if greater than or equal to 5% of the mix and the sum of all screened nuclides cannot exceed 10%. In its letter dated May 30, 2024, the NRC staff asked the licensee to clarify establishing a DCGLSURROGATE for Ni-63, which has been dispositioned as an insignificant contributor even though it exceeds 5% of the dose in the mixture based on Table 6-21, Relative Dose Fractions. In its RAI response dated June 27, 2024, the licensee agreed in response RAI 5-2, that Ni-63 should be accounted for by using Co-60 as a surrogate. As such, there are three ROCs for the FSS (Cs-137, Co-60, and Ni-63).

5.2.2 Insignificant Contributors Section 6.13, of the LTP discusses the methodology for determining the Insignificant Dose Contributors (IDCs) and the licensees anticipated ROCs after the IDCs are deselected. The IDCs are determined by calculation of the Relative Dose Fraction (RDF). The RDFi,k, for radionuclide, i, in each sample is calculated using the radionuclide fractions depicted in the LTP, specifically, Table 6-4 of the LTP, the applicable DCGLs from Table 6-17, Surface Contamination Limits (DCGLs), and Equation 6-5. Table 6-21, Relative Dose Fraction, in the LTP contains the RDFs for each radionuclide in each of the 6 samples (5 composite smears and one sludge sample) that were sent for HTD analysis. The RDFs for Co-60 and Cs-137 were summed for each sample. The RDFs for the remaining radionuclides were summed for each sample.

The licensee identified only three ROCs, Ni-63, Cs-137 and Co-60. The dose for the other radionuclides in five out of the six samples is less than 1 percent. The sixth sample is 8 percent of the dose. The results of the calculations are presented in Table 6-22, Radionuclides of Concern (ROCs) and Insignificant Dose Contributors (IDCs), in the LTP. The licensee further states that, because this evaluation was based on a dose limit of 15 mrem/yr, no adjustment (reduction) in the DCGLs is necessary or will be performed. Dose contribution from the insignificant contributors has to be considered when demonstrating compliance with the dose based release criteria. However, the NRC staff determined that because the licensee is proposing use of DCGLs corresponding to 15 mrem/yr, the dose contribution from the insignificant contributors would not be significant nor impact the ability to demonstrate compliance with the NRCs dose criteria of 25 mrem/yr.

5.2.3 NRC Evaluation of ROC Determination The NRC staff evaluated the licensees proposed ROCs, insignificant contributors, and use of surrogates in accordance with the regulatory guidance and acceptance criteria contained in NUREG-1757, Volume 2, Revision 2, Appendix I, Technical Basis for Site-Specific Dose Modeling Evaluations, and Section 2.5 of NUREG-1700. The NRC staff finds the licensees use of decay corrected activity and the 75th percentile values to determine mixture fractions and

45 insignificant contributors as well as using 95th percentile values to establish surrogate ratios to be reasonable.

The licensee evaluated 11 radionuclides and determined all but three (Ni-63, Cs-137 and Co-60) were IDCs. The mixture fractions are based on laboratory analysis of five composite samples of smears taken in various areas of the NS Savannah and one sludge sample from the makeup storage tank. While the licensee did not evaluate the smear composites for certain hard-to-detect radionuclides and the averaging method performed then significantly reduced the fractional makeup of the same radionuclides, the licensee ultimately evaluated the potential dose contributions on a sample by sample basis to demonstrate that the fractional make up of each sample resulted in only Co-60 and Cs-137 contributing over 90 percent of the potential dose.

The NRC staff reviewed the determinations and finds the assessment performed by the licensee provides reasonable assurance that Cs-137 and Co-60 are properly considered ROCs for the FSSs. The NRC staff notes that Ni-63 is also considered a ROC due to not meeting the criteria for being screened out in Section 5.4.5.2 of the LTP as it exceeds 5% of the total activity in all samples evaluated. The remaining radionuclides, in aggregate, would contribute less than 10%

of the dose. This is consistent with guidance in Section 3.3, Insignificant Radionuclides and Exposure Pathways, of NUREG-1757, Volume 2, Revision 2, and MARSSIM Section 4.3.2, DCGLs and the Use of Surrogate Measurements.

5.3 Site Release Criteria Section 6.2, Proposed Radiological Criteria for License Termination, of the LTP acknowledges the dose based license termination regulation at 10 CFR 20.1402 which provides a dose criterion of 25 mrem/yr and also being ALARA. The licensee elaborates that it has adopted an administrative dose standard of 15 mrem/yr and explains the factors influencing this decision.

5.3.1 DCGLs and Dose Factors Section 6.10, Analysis and Results, of the LTP provides the results of simulations that were run for the Remediation and Component Removal worker scenarios. Table 6-17, Surface Contamination Limits (DCGLs), in the LTP provides the surface DCGLs corresponding to 15 mrem/yr that the licensee is proposing for the NS Savannah. The NRC staff edited this table to only address the ROCs the licensee has proposed as shown below. The DCGLs only address surface contamination as the licensee has stated that there is no volumetric contamination present and, being a ship, there is no groundwater to be concerned with.

Table 6-17 (modified) Structural Surface DCGLs corresponding to 15 mrem/yr Radionuclide DCGL (dpm/100cm2)

Ni-63 2.53E+08 Co-60 2.37E+04 Cs-137 1.20E+05 Section 5.4.5.3 of the LTP discusses the use of area factors and the DCGLEMC for the NS Savannah. The licensee states that a review of the scanning MDCs for the instruments to be

46 used for the FSS show that MDCs are a small fraction of the DCGLs, so no area factors have been calculated and the DCGLEMC will not be used.

Section 5.7.1.4 Samples, of the LTP, notes that, if water or sludge is encountered in a system during FSS, sample results will be compared to the Effluent Concentrations listed in Table 2, Column 2 of Appendix B to 10 CFR 20. If the sample results are greater than the Effluent Concentrations, the medium will be remediated or removed. In its letter dated September 16, 2024, the NRC staff asked the licensee to provide a discussion of how residual radioactivity from water and/or sludge remaining at license termination will demonstrate compliance with 25 mrem/yr. In its RAI response dated October 16, 2024, the licensee stated in response RAI 5-6 that, at license termination, no reactor plant-related water or sediment will remain on the ship.

5.3.2 ALARA Evaluation NRC guidance for conducting ALARA analyses is provided in NUREG 1757, Volume 2, Revision 2. In Section 4.4, ALARA Evaluation, of the LTP, the licensee provides its ALARA analysis process and conclusions for the remaining components. The NRC staff notes that presumptions in this section, including the licensees stated FSS goal to meet a self-imposed concentration limit of 15mrem/yr, differ from NRC guidance. The NRC staff evaluated the LTP to verify that the licensee satisfied the requirement that the plans described in the LTP provide reasonable assurance that the licensee will be able to perform adequate surveys to demonstrate compliance with the radiological criteria for unrestricted use, as specified in 10 CFR 20.1402.

The NS Savannah project is nearing completion and, based on the conservative decommissioning activities, a pre-determined compliance measure for ALARA is not suitable for this vessel. The licensee has chosen to commit to a performance-based compliance method for ALARA similar to that mentioned in Section 6.3.6.2 of NUREG-1757, Volume 2, Revision 2. In its letter dated December 16, 2024, the NRC staff asked the licensee to confirm the NRC staffs understanding that key ALARA actions (see below) have been, and would be, in place and are demonstrable for ALARA.

The NRC staff asked for confirmation on the accuracy of the following actions which will ensure that actual residual radioactivity remaining at license termination will coincide with a much lower hypothetical dose to an average member of the critical group than 25 mrem/y and will demonstrate the residual radioactivity in the NS Savannah at license termination is ALARA:

A radiation safety officer (RSO) position has been and will be maintained until license termination. Throughout the remediation process, the RSO routinely evaluated work activities, contamination levels, and worker exposures to ensure ALARA was maintained. The RSO will continue to perform such activities as required until the license is terminated.

An administrative limit of 15 mrem/y was established to ensure the actual residual radioactivity will be significantly less than 25 mrem/y.

No elevations of residual radioactivity will exceed the 15 mrem/y derived concentration guideline limit (DCGLw) concentrations.

All sediment (sludge) and water containing residual radioactivity will be removed.

Processes are in place that are more restrictive than in the license termination plan (e.g.,

processes or procedures are established for decontamination personnel to attempt to remediate any area causing instruments to have readings in excess of 1,000 cpm);

47 Removable activity will be reduced to less than 10 percent of the total activity present consistent with assumptions associated with developing the DCGLs.

In its response dated December 19, 2024, the licensee confirmed the accuracy of the information above and that these actions will ensure that actual residual radioactivity remaining at license termination will coincide with a much lower hypothetical dose to an average member of the critical group than 25 mrem/y and will demonstrate the residual radioactivity in the NS Savannah at license termination is ALARA. The NRC staffs evaluation of the licensees proposed ALARA process can be found below, in Section 5.3.5 of this SER.

5.3.3 DCGLs for Surrogates and Gross Activity Measurements The licensee indicated that Cs-137, Co-60, and Ni-63 will be the ROCs for NS Savannah. Of these, Ni-63 is a hard-to-detect radionuclide in that its radioactive emissions are not readily detectable by hand-held instruments because it is a low energy beta emitter. Ni-63 composes greater than 5% of the activity present in samples collected to evaluate the activity ratios so it cannot be considered an insignificant contributor. Because much of the surveys are anticipated to be performed using hand-held instruments that measure gross beta/gamma activity, the DCGLs need to be adjusted to account for using surrogate radionuclides (Co-60 will be used as a surrogate for Ni-63) and gross activity measurements.

Sections 5.4.5, Selection of DCGLs, and 5.4.5.1, Gross Activity DCGLs, of the LTP states that the radionuclide specific DCGLs will be used to establish gross activity DCGLs. These gross activity DCGLs will be established based on a representative radionuclide mix established for the entire ship. In cases where measurable activity still exists, scaling factors will be used to establish the activity contribution for any hard-to-detect radionuclides that may be present.

Scaling factors will be selected from available composite waste stream analyses or similar assays. Such analyses may be performed periodically and documented in support of waste characterization needs.

For cases of survey units for which there is no measurable activity distinguishable from background, a representative radionuclide mix may be selected based upon historical characterization information for the survey unit of interest or for units with similar history and physical characteristics (e.g., information from adjacent areas). Gross activity DCGLs will be established for gross beta measurements using equation 5-1 in the LTP (copied below). Only gross beta/gamma measurements are anticipated; no gross alpha activity measurements will be made. Equation 5-1 will be applied to all ROCs. Hard-To-Detect radionuclides (e.g., Ni-63) will be included by surrogate ratio DCGL as previously discussed.

Section 5.4.5.2 of the LTP discusses surrogate measurements. In its letter dated May 30, 2024, the NRC staff asked the licensee to clarify establishing a DCGLSURROGATE for Ni-63. In its RAI response dated June 27, 2024, the licensee agreed in response RAI 5-2, that it would account for Ni-63 activity by adjusting the Co-60 DCGL as a surrogate measurement. The licensee used equation 5-2 in the LTP (copied below) and the Ni-63/Co-60 ratio of 55.3 (based on activity fractions in Table 6-5, Radionuclide Sum of Fractions and Normalized Sum of Fractions, in the LTP).

48 Because the DCGL for Ni-63 is much greater than that for Co-60, the resulting adjusted DCGL for Co-60 is only slightly less than the unadjusted DCGL (2.36E4 dpm/100cm2 vs 2.37E4 dpm/100cm2).

5.3.4 Background and Reference Areas/Measurements Section 5.2 of the LTP states that background radiation in the NS Savannah, being isolated from marine and terrestrial environments, is anticipated to be lower than typical sites. If any ROCs are present in background, planning efforts may include establishing appropriate reference areas and reference materials.

In its letter dated September 16, 2024, the NRC staff asked the licensee to clarify how ambient background will be appropriately/conservatively determined for each gross activity measurement. In its RAI response dated October 16, 2024, the licensee stated in RAI 5-11 that FSS data will only be evaluated using the Sign Test because the ROCs do not exist in background or are not present in a concentration that is a relevant fraction of the DCGLW. For this reason, the licensee only anticipates subtraction of ambient beta/gamma from their static and scan measurements. The licensee will utilize gross ambient beta/gamma measurements when demonstrating compliance with the DCGLs.

5.3.5 NRC Evaluation of Site Release Criteria The NRC staff evaluated the administrative level DCGLs corresponding to 15 mrem/yr on surface structures in Section 6 of this SER and considers it adequate. While the licensee has presented a self-imposed administrative limit of 15 mrem/y, the NRC staff will consider compliance to be the demonstration of meeting a 25 mrem/yr limit. The licensee has developed no area factors for elevated measurement comparisons, and has stated in response to RAI 5-3 that it would not have any exceedances of the DCGLw at license termination; therefore, if the licensee does not meet the DCGLWs for 25 mrem/yr, a request for an amendment to the license may be needed.

As discussed in Section 5.3.2 of this SER, the licensees plans to reduce levels of residual radioactivity to a self-imposed limit of 15 mrem/yr differs from NRC guidance on pre-determined ALARA compliance measures, found in Section 6.3.6.2 of NUREG-1757, Volume 2, Revision 2.

However, the NRC staff evaluated the actions and commitments throughout the LTP, verified in the letter dated December 19, 2024, and determined they are adequate to ensure ALARA is maintained through license termination consistent with the requirements of 10 CFR 20.1402.

The NRC staff verified the licensees derivation of surrogate DCGLs (Co-60 is the proposed surrogate for Ni-63) and finds the calculation accurate. Similarly, the equation for derivation of the gross DCGLs is accurate. The methods proposed are consistent with NUREG-1575.

Because the licensee is proposing to only perform surface structure measurements/scanning and to utilize the Sign Test for demonstrating compliance with the DCGLs, it is only planning to consider gross ambient beta/gamma measurements. As such, there are no concerns that the appropriate background materials are not being considered. This was clarified in its RAI 5-11 response dated October 16, 2024, in which the licensee states that it will be ignoring ambient

49 background and conservatively eating background in its measurements to demonstrate compliance with the DCGLs.

Given all of the above, the NRC staff finds the release criteria proposed by the licensee to be reasonable and ALARA and sufficient to demonstrate, in part, compliance with 10 CFR 50.82(a)(9)(ii)(C).

5.4 FSS Implementation Section 5.4.4, Area Preparation: Isolation and Control of the LTP discusses what must be done to transition an area to be ready for FSS. A walkdown will occur to establish if the area is ready for final survey activities and identify any work practice issues that must be addressed in survey planning and design. Determination of readiness for FSS will be based on a characterization survey, turnover survey and/or a RASS indicating that the residual radioactive material is likely to comply with the FSS criteria. Isolation and control criteria are presented in this section and, once the area meets the criteria, isolation and control will be achieved through a combination of personnel training, physical barriers, postings, and site notices, as appropriate, to prevent unauthorized access to an isolated survey unit. Isolation and control measures will be implemented through approved plant procedures. An administrative process will be used to evaluate, approve (or deny), and document all activities conducted in these areas during and following FSSs.

5.4.1 Survey Methods Section 5.7.1, Survey Methods, of the LTP and its subsections state that scanning is generally described as the process of moving portable radiation detectors across a surface with the intent of locating residual radioactivity. Section 5.7.1.2, Fixed Measurements, of the LTP, states that these measurements will be taken by placing the instrument at the appropriate distance above the surface, taking a discrete measurement for a pre-determined time interval, and recording the reading. Fixed measurements may be taken as part of investigations of action level exceedances, professional judgment, or at locations established by a given survey design.

Sections 5.1, Introduction, and 5.7.1.3, Advanced Technologies, of the LTP state that the licensee does not expect to use advanced technologies for survey methods. Section 5.5, Final Status Survey Design Elements, in the LTP states that other advanced survey methods will be used for the FSS for the NS Savannah rooms and Class 1 and Class 2 areas. Thus, there are conflicting statements in the LTP regarding the use of advanced technology survey methods.

In its letter dated May 30, 2024, the NRC staff asked the licensee to clarify whether advanced technologies, not found in NRCs guidance, will be used for surveys and, if so, describe what technology is anticipated, how it will be utilized, and what will be the sensitivity of the technology for the ROCs. In its RAI response dated June 27, 2024, the licensee stated in response RAI 5-4 the following:

[Advanced technologies] refer to survey instruments or methods that do not use conventional hand-held detectors connected to ratemeters or scalers. The ISOCS has been widely used for final status surveys at recent nuclear power plant decommissioning projects, primarily for subsurface concrete evaluations. MARAD is planning on using the ISOCS for measurements on the steam drums and potentially on portions of the CV, RC or other areas where ISOCS measurements can save time and cost rather than using conventional survey

50 equipment. The count times will be long enough to ensure that the MDCs for any count will fall in the MARSSIM recommended range of 10 to 50 percent of the DCGLs.

Section 5.6, Survey Protocol for Non-Structural Systems and Components, in the LTP states that surface activity assessments for non-structural systems and components can be made by making measurements at traps, tanks, open piping and other appropriate access points where activity levels should be representative of those on the interior surfaces. Assessments may also be made via in-situ gamma spectroscopy or pipe crawlers, provided adequate instrument efficiencies and detection limits can be achieved. Detection limits for surface activity assessments should be at least 50% of the release limits.

Section 5.7.1.4 of the LTP provides that sampling is the process of collecting a portion of a medium as a representation of the locally remaining medium. The collected portion of the medium is then analyzed to determine the radionuclide concentration. Examples of materials that may be sampled include trap sediments and lagging. Samples will be collected during characterization surveys. No samples are anticipated during FSSs.

5.4.2 Instrumentation Section 5.7.2.1, Instrument Selection, of the LTP states that the survey instrument must be able to detect the type of radiation of interest, and, depending on the application, the measurement system should be capable of measuring levels that are less than the DCGLW.

However, in some cases instruments used for scanning may have detection limits that are greater than the DCGLW. This is allowed by MARSSIM and is acceptable as long as the grid spacing (for Class 1 survey units) and investigation levels used are in accordance with this plan.

Instruments that the licensee is planning to utilize for the FSSP are listed in Table 5-4, Available Instruments and Associated MDCs, in the LTP along with the nominal fixed measurement MDC and Scan MDC. The scanning MDCs for the instruments to be used for the FSS are a small fraction of the DCGLs. Because the scan MDC for the instruments appears to be significantly less than the DCGLw the licensee has proposed, no modification to the number of samples determined statistically under MARSSIM guidance is anticipated.

The LTP describes that the instruments used for fixed measurements to demonstrate that the average concentration in a survey unit is less than the DCGLW should have an MDC that is no greater than 50% of the DCGLW. However, though it is not anticipated, there may be special circumstances where the best available technology cannot meet this goal. In such a case, measurements will be made at the best MDC that can be achieved. As previously mentioned, the instruments that will be utilized to perform the FSS are presented in Table 5-4 of the LTP along with the fixed measurement nominal MDC for either a 1-minute or 10-minute scaler count time depending on the instrument.

The NRC staff notes that Table 5-4 does not indicate if source efficiency corrections were included in the nominal fixed and scan MDC values. In addition, source efficiency values for maximum beta energies between 150-400 keV are not indicated. In its letter dated September 16, 2024, the NRC staff asked the licensee if the nominal MDC and nominal scan MDC values in Table 5-4 include source efficiency correction. In its RAI response dated October 16, 2024, the licensee stated in response RAI 5-5 that Table 5-4 will be revised to list the MDC for the ISOCS system in dpm/100 cm2. In all cases, the presented nominal MDC is at most ~10% of the proposed DCGLw.

51 As was previously mentioned regarding the licensees response to RAI 5-4, the licensee is planning on using the ISOCS for measurements on the steam drums and potentially on portions of the CV, RC or other areas where ISOCS measurements can save time and cost rather than using conventional survey equipment. The count times will be long enough to ensure that the MDCs for any count will fall in the MARSSIM recommended range of 10 to 50 percent of the DCGLs. The nominal MDCs presented in Table 5-4 for Cs-137 and Co-60 for the HPGe detector are ~2,200 dpm/100cm2 for a 1000 second count.

5.4.3 Quality Assurance Commitments for the FSS Section 5.7.2.2, Calibration and Maintenance, of the LTP states that all instrumentation used for measurements to demonstrate compliance with the radiological criterion for license termination at the NS Savannah will be calibrated and maintained under approved procedures and the DQAP or vendor QA plan that satisfies the requirement of the DQAP. Instruments will be calibrated for normal use under typical field conditions. Calibration standards will be traceable to the National Institute of Standards and Technology. Instruments used to measure gross beta surface activity will be calibrated to Tc-99 or Co-60 to bound the beta energies for the beta-emitting radionuclides that will be encountered during final survey activities. Instrument efficiencies may require modifications to account for surface conditions or paint coverings. Such modifications, if necessary, will be established using the information in Section 5 of NUREG-1507 and pertinent site characterization data.

In its letter dated September 16, 2024, the NRC staff asked the licensee to clarify that operational checks of survey instruments are being conducted daily or per shift when used. In its RAI response dated October 16, 2024, the licensee stated in response RAI 5-9 that instrumentation will be checked for proper operation at least daily, in accordance with approved procedures. If the instrument operational check does not fall within the established range, the instrument will be removed from use until the reason for the deviation can be resolved and acceptable operation is again demonstrated. If the instrument fails a post-survey source check, all data collected during that time period with the instrument will be carefully reviewed and possibly adjusted or discarded, depending on the cause of the failure.

Section 5.11, Quality Assurance and Quality Control Measures, of the LTP states that QA and QC measures are integrated into all decommissioning activities, including implementation of the FSS. All FSS activities essential to data quality will be implemented and performed under approved procedures. Effective implementation of administrative controls will be verified through self-assessments, monitoring and audit activities, with corrective actions being prescribed, implemented and verified in the event any deficiencies are identified. These measures apply to the related services provided by off-site vendors. Note that self-assessments are performed by individuals with direct responsibilities in the area they assess. Audits and monitoring are performed by individuals with no direct responsibilities in the area they are auditing or monitoring.

52

53 5.4.4 NRC Evaluation of FSS Design and Measurement Quality The NRC staff evaluated the licensees commitments for survey methods, instrumentation, and QA requirements with respect to the information contained in NUREG-1757, Volume 2, Revision 2. In general, the NRC staff finds the commitments to be very general with respect to the methods to be employed. The NRC staff verified the licensees estimates for scan MDC for the pancake Geiger-Mueller (GM) probe of <8,000 dpm/100 cm2 using similar detector and background characteristics as presented in Table 5-4 in the LTP.

However, the NRC staff noted that this estimate did not include the source efficiency term and also the d value of 1.38 is not adequately justified. When the NRC staff incorporated these checks, the scan MDC for this particular instrument increased to levels greater than the DCGLw. The NRC staff made a similar effort to check the other instruments and fixed measurement MDC and well as the scan MDC. In both cases, the NRC staff could generally verify the calculations performed by the licensee but noted discrepancies in the source efficiencies and d value selected. While the scan MDCs significantly increased, only the pancake GM probe had a scan MDC above 50% of the proposed DCGL which, consistent with guidance, makes it unsuitable for use as a FSS instrument.

In its RAI 5-5 response dated October 16, 2024, the licensee revised Table 5-4 in the LTP and footnoted the basis for the stated nominal MDCs. This updated table is reflected above, as Table 5-4 in this SER. The NRC staff reviewed the revised information as well as CR-164, Calculation of Weighted Efficiencies and MDCs of Ludlum Detectors for Final Status Surveys on the NS SAVANNAH, and found it adequate for surveys of the ROCs and measurement types that are to be taken to demonstrate compliance with the license termination criteria. With these modifications to the LTP, the NRC staff finds the LTPs commitments for survey implementation are consistent with NUREG-1575, Revision 1, and NUREG-1757, Volume 2, Revision 2; and demonstrate, in part, compliance with 10 CFR 50.82(a)(9)(ii)(C).

5.5 FSS Assessment and Report Section 5.5, Final Status Survey Design Elements, of the LTP states that the general approach prescribed by MARSSIM for FSSs requires that at least some minimum number of measurements or samples be taken within a survey unit, so that the non-parametric statistical tests used for data assessment can be applied with adequate confidence. Decisions regarding whether a given survey unit meets the applicable release criterion are made based on the results of these tests. Scanning measurements are used to check the design basis for the survey by evaluating if any small areas of elevated activity exist that would require re-classification, tighter grid spacing for the fixed measurements, or both.

Section 5.8, Survey Data Assessment, of the LTP discusses assessment of FSS data including verification that QA/QC procedures were used during data acquisition. Prior to evaluating the data collected from a survey unit against the release criterion, the data are first confirmed to have been acquired in accordance with all applicable procedures and QA/QC requirements. Any discrepancies between the data quality or the data collection process and the applicable requirements are resolved and documented prior to proceeding with data analysis.

Data assessment will be performed by trained personnel using approved procedures.

The LTP further provides that the first step in the data assessment process is to convert all survey results to units consistent with the DCGL values. Next, the individual measurements and

54 sample concentrations will be compared to DCGL levels for evidence of small areas of elevated activity or results that are outliers to the rest of the measurements.

In its letter dated May 30, 2024, the NRC staff asked the licensee to confirm plans to remediate materials with measurements above the DCGLW until residual radioactivity is less than the DCGLW or provide instructions on how any elevated measurements will be addressed in demonstrating compliance with 10 CFR 20.1402. In its RAI response dated June 27, 2024, the licensee committed in response RAI 5-3 to remediating or removing all exceedances of the DCGLW.

As a result, the NRC staff determines that there is no expected need for DCGLemc. Graphical analyses of survey data that depict the spatial correlation of the measurements are especially useful for such assessments and will be used to the extent practical. The results may indicate that additional data or additional remediation and re-survey may be necessary. If this is not the case, the survey results will then be evaluated using direct comparisons or statistical methods, as appropriate, to determine if they exceed the release criterion. If the release criterion has been exceeded or if results indicate the need for additional data points, appropriate further actions will then be determined.

The initial data evaluation will be as described in Table 5-5 of the LTP, copied and pasted below.

Sections 5.8.1, Sign Test, and 5.8.2, Unity Rule, of the LTP provide that when radionuclide specific measurements are made in survey units having multiple radionuclides, compliance with the radiological release criterion will be assessed through use of the unity rule, also known as the sum of fractions. The unity rule, represented in equation 5-9 of the LTP which is copied and pasted below, is satisfied when radionuclide mixtures yield a combined fractional concentration limit that is less than or equal to one.

55 The licensee notes that given the site contains no soil, laboratory analysis is not anticipated.

Similarly, split sampling is not anticipated given that it is typically associated with evaluating soil.

The licensee states that FSS data will only be evaluated with the Sign Test because the radionuclides of interest either do not exist in background or are not present in a concentration that is a relevant fraction of the DCGLW. When radionuclide specific measurements are made in survey units having multiple radionuclides, compliance with the radiological release criterion will be assessed through use of the unity rule, also known as the sum of fractions. This is shown in equation 5-9, copied and pasted above.

5.5.1 FSS Report Section 5.10, Final Status Survey Release Records and Reports, of the LTP discusses what will be included in the FSS report.

The documentation describing the FSS for a given survey unit will include:

A physical description of the survey unit; A summary of any characterization data associated with the survey unit including any required investigations, re-classifications or subdivisions; The classification history of the unit; A description of remediation activities (if any) performed during FSS; Results and discussion of any ALARA evaluations, if performed; A discussion of the survey design (combination of scans and fixed measurements used; number of measurements; grid spacing; etc.);

Tabular and graphical depictions of survey results including QC results; Discussions of data assessments; and, A statement that the survey unit meets the applicable release criteria.

The LTP further states that the FSS results will be documented and made available to the NRC for multiple survey areas rather than for individual survey units. Reports will be compiled after FSS activities for all the survey units within a given area are completed. These reports will be prepared and submitted per NS Savannah Administrative Procedures.

In its letter dated September 16, 2024, the NRC staff asked the licensee to clarify the actions that will be taken if a survey unit fails. In its RAI response dated October 16, 2024, the licensee stated in response RAI 5-10, that if a survey unit fails, the cause will be evaluated, and corrections will be performed. This may also include additional remediation and resurveying.

These actions are defined in the licensees procedure STS-005-033, Final Status Survey Data Assessment and Investigation. After all actions have been taken and completion of a successful FSS, the FSS Release Record will include a discussion of the failure, any follow-up actions and the results of the survey data and data analysis from the successful survey.

5.5.2 NRC Evaluation of FSS Assessment and Report The NRC staff evaluated the licensees statements that it will be remediating or removing any contamination which exceeds the DCGLW. The licensee indicated that it will conduct scanning and surveys consistent with MARSSIM guidance measuring the gross beta/gamma emissions on structural surfaces and components and comparing the measurements to the DCGLs previously mentioned. As one ROC (Ni-63) is a hard-to-detect radionuclide, it will utilize Co-60 as a surrogate for that radionuclide and modify that DCGL consistent with MARSSIM guidance.

56 All measurements, whether it be of structural surfaces or components will be evaluated for each survey unit using the Sign Test. A final status survey report of the various FSSs will be submitted once all the individual surveys in a survey area are completed. The NRC staff finds that the preceding sections in Chapter 5 demonstrate compliance with 10 CFR 50(a)(9)(ii)(C) and, if properly implemented and barring unforeseen circumstances, should be capable of demonstrating compliance with applicable requirements in 10 CFR 20.1402 and 10 CFR 20.1501.

6 COMPLIANCE WITH RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION Subpart E of 10 CFR Part 20, Radiological Criteria for License Termination, establishes criteria for the release of sites for unrestricted use. Specifically, per 10 CFR 20.1402, the residual radioactivity that is distinguishable from background levels must result in a TEDE to the average member of the critical group (AMCG) that does not exceed 0.25 mSv/yr (25 mrem/yr) and the residual radioactivity must also be reduced to levels that are ALARA. Details related to the application of ALARA are discussed in Chapters 4 and 5 of the LTP. Chapter 6, Compliance with the Radiological Criteria for License Termination, of the LTP describes the dose modeling and calculations used to establish the site-specific DCGLs that the licensee will apply to the NS Savannah site during final status surveys in order to demonstrate compliance with the radiological criteria for release for unrestricted use contained in 10 CFR 20.1402. The NRC staff reviewed this information using Section 2.6, Compliance with the Radiological Criteria for License Termination, of NUREG-1700, which refers to multiple sections in NUREG-1757, Volume 2, Revision 2, for additional details. In addition, the NRC staff is assessing the licensees compliance with the requirements set forth in 10 CFR 50.82(a)(9)(ii)(D), which requires the LTP to include detailed plans for the final radiation survey, because the DCGLs are used in the final radiation surveys. As noted in Section 1.3 of this SER, the NRC staffs review of the dose assessment and DCGL development is predicated on the site conditions being as described in the LTP.

6.1 Approach for Overall Dose Compliance As the NS Savannah is not a typical land-based site, the only potentially contaminated media expected to remain onsite at the time of license termination are ship surfaces. As Section 6.1, Introduction, of the LTP describes, the licensee has provided dose analyses over a 70-year period, consistent with the assumed lifespan of buildings following license termination guidance in NUREG-1496, Generic Environmental Impact Statement in Support of Rulemaking on Radiological Criteria for License Termination of NRC-Licensed Nuclear Facilities, dated July 1997 (ML042310492). To demonstrate that the overall combined dose at the site from all sources at the time of license termination is consistent with the criteria for unrestricted use in 10 CFR 20.1402, the licensee will apply the following equation, found within Section 6.10 of the LTP:

=

Equation 6-4 in the LTP Where:

= Sum of the fractions for all radionuclides of concern

= Surface Activity Concentration of radionuclide, 1002

= Derived Concentration Guideline Level of radionuclide, 1002

57 Consistent with NRC guidance in Section 2.7, Sum of Fractions, NUREG-1757 Volume 2, Revision 2, the sum of fractions (SOF) approach in the LTP is an NRC-approved approach to account for dose from multiple radionuclides in order to demonstrate compliance with the 0.25 mSv/yr (25 mrem/yr) TEDE criterion in 10 CFR 20.1402. The sum of the relative ratios for all the ROCs may not exceed 1.

As described in NUREG-1757, Volume 2 Revision 2, the use of target values or DCGLs is a practical method to develop reasonable surveys (i.e.,10 CFR 20.1501(a) and (b)) and to evaluate the dose consequence of survey results (i.e., 10 CFR 20.1402).

6.1.1 Methods for Evaluating Dose and Establishing DCGLs The methods for evaluating the doses and SOFs from the source term are described in detail in Chapter 6 of the LTP. The licensee developed site-specific DCGLs to evaluate the potential dose from residual radioactivity on surfaces at the site. Because the NS Savannah has been and is expected to remain isolated from land through license termination, the licensee is not considering potential contamination from soil, buried structures, surface water, or groundwater.

The licensee evaluated multiple ROCs for multiple plausible exposure scenarios, then selected most limiting DCGL for each ROC. DCGLs have been established for the three expected radionuclides of concern (ROC), Cs-137, Co-60, and Ni-63, for surfaces that correlate to an adopted dose standard of 0.15 mSv/yr (15 mrem/yr), in part as an ALARA practice.

The licensee expects the ship to remain intact and does not expect to dismantle the ship prior to license termination. However, the licensee has elevated multiple exposure scenarios for possible end-state conditions to evaluate dose, including scenarios for offsite scrapping of the ship. The licensee developed site-specific exposure scenarios and referenced NRC guidance established in NUREG-1640, Radiological Assessments for Clearance of Materials from Nuclear Facilities: Main Report, dated June 2003 (ML031700258), to evaluate dose in the event of two plausible end-state conditions, preservation and shipbreaking, and one less likely but plausible end-state condition, artificial reefing. Each potential end-state is modeled using various exposure scenarios which are discussed in more detail in Section 6.3 of this SER.

6.1.2 NRC Evaluation of Approach for Overall Dose Compliance The NRC staff has reviewed and evaluated the approach and equation in the LTP and finds that the approach considers expected ROCs from all source terms (i.e., potentially contaminated media) onsite. The NRC staff finds the licensees approach for accounting for multiple ROCs using the SOF approach is consistent with NRC guidance in NUREG-1575 and NUREG-1757, Volume 2, Revision 2.

The NRC staff has reviewed the licensees approach to demonstrate compliance with a licensee-adopted dose standard of 0.15 mSv/yr (15 mrem/yr). As previously stated, NRC staff has evaluated the licensees proposed LTP, including its associated dose assessments, against the 25 mrem/yr dose limit in 10 CFR 20.1402. As such, this SER will discuss the NRC staffs findings and evaluations of the LTP based on the dose criterion of 0.25 mSv/yr (25 mrem/yr) in 10 CFR 20.1402.

The NRC staff finds the approach to using the exposure scenarios correlating to ship preservation, shipbreaking, and artificial reefing acceptable for determining dose to an AMCG.

58 The NRC staff notes that the licensee has evaluated twenty (20) exposure scenarios and conducted detailed technical analysis of fifteen (15) exposure scenarios. The licensee has selected the most conservative dose estimate for each ROC. Furthermore, the licensee has followed NRC guidance in NUREG-1640 and NUREG-1757, Volume 2, Revision 2, and provided dose models that incorporate both realistic and conservative parameters.

In its letter dated September 16, 2024, the NRC staff asked the licensee to explain the source terms that have been modeled and their relevance to the residual radioactivity that may be present at license termination. In its RAI response dated October 16, 2024, the licensee stated in response RAI 5-6 that no reactor plant-related water or sediment will remain on the ship.

In its letter dated September 16, 2024, the NRC staff asked the licensee to provide any updates to the end-state configuration plans of any pipe or systems to remain at license termination, including plans to fill any pipe or system to make [it] inaccessible. In its RAI response dated October 16, 2024, the licensee confirmed in response RAI 5-7 that both structural and non-structural systems will be considered for the total compliance dose equation. Because FSS plans for both include taking smears and samples of various surfaces, the FSS results will be compared directly and applied to the SOF equation (Equation 6-4).

Therefore, the NRC staff is approving this LTP on the assumption that the TEDE to a receptor will include both structural and non-structural systems and will not include reactor plant-related water or sediment. For these reasons, the NRC staff concludes that the compliance equations in the LTP are acceptable for demonstrating that the final combined site dose is less than the unrestricted release criteria in 10 CFR 20.1402.

6.2 Exposure Scenarios, Critical Group, and Pathways Chapter 6 of the LTP presents the exposure scenarios the licensee evaluated correlating to ship preservation, shipbreaking, and artificial reefing end-state conditions. The licensee has evaluated twenty (20) possible exposure scenarios and conducted detailed technical analysis of fifteen (15) plausible exposure scenarios. The licensee has selected the most conservative dose estimate for each ROC to be the basis of their DCGLs.

6.2.1 Exposure Scenarios, Critical Group, and Pathways At the time of license termination, the NS Savannah is expected to be intact. Upon release for unrestricted use, the licensee can disposition the ship in one of three ways based on its programmatic agreement: physical destruction and recycling of the ship through shipbreaking; preservation for public use; and beneficial reuse by sinking the vessel in shallow water to form, or act as part of an existing, artificial reef.

As described in Section 6 of the LTP, shipbreaking and preservation are considered reasonably foreseeable end-state conditions. Artificial reefing has been designated as a less-likely but plausible exposure scenario because the licensee policy prohibits ships constructed before 1985 from the artificial reef program. Furthermore, of the 231 ships the licensee disposed between 2010 and 2020, only four (4) of them have been reefed. The shipbreaking and artificial reefing conditions would require similar dismantlement activities, and therefore the licensees analysis adequately considers removal of structures and components in both plausible and less-like but plausible scenarios.

59 The licensee evaluated sixteen (16) exposure scenarios related to the shipbreaking end-state condition and four (4) exposure scenarios related to the preservation end-state condition. Table 6-1 below provides an overview of the exposure scenarios evaluated for shipbreaking and preservation and the AMCG in each scenario. This table was adapted from Table 6-6 and additional text in Chapter 6 of the LTP.

Table 6-1: Exposure Scenarios Evaluated for Calculation of Surface DCGLs for Preservation and Shipbreaking End-State Conditions Ship Status Description Exposed Individual Significance Preservation Office Worker/

Tour Guide*

Adult worker Significant:

Full-time employee Preservation Housekeeping Adult worker Insignificant: part-time employee Preservation Minor repairs/maintenance Adult worker Insignificant: part-time employee Preservation Tours and Meetings on ship Members of the Public Insignificant: few hours per year Pre-Shipbreaking Housekeeping Adult worker Insignificant: part-time employee Pre-Shipbreaking Minor repairs/Maintenance Adult worker Insignificant: part-time employee Shipbreaking Remediation of hazardous materials on ship*

Adult worker Significant:

Full-time employee Shipbreaking Component removal/metal cutting on ship*

Adult worker Significant:

Full-time employee Shipbreaking 7 steel handling and processing scenarios defined in NUREG-1640*

Adult worker Significant:

Full-time employee Dismantlement complete 5 groundwater infiltration by leachate scenarios from landfills or storage piles defined in NUREG-1640*

Members of the Public Significant:

Potential daily exposure

• Detailed technical analysis performed due to risk significance The licensee designated fifteen (15) exposure scenarios as significant, and the licensee performed detailed technical analysis which is summarized in the subsections below. Dose to the AMCG that was designated as significant consider those who are fully time employees at the task, or their tasks require handling or manipulating potentially contaminated components.

6.2.2 Preservation Exposure Scenarios Section 6.9, Inputs to the Scenario Calculations, of the LTP designated the Office Worker/Tour Guide, herein referred to as Tour Guide scenario as significant. In this exposure scenario, the AMCG, a tour guide is assumed to work on the ship full-time (eight hours per day, 250 days per year, 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year) in the Containment Vessel and Reactor Compartment.

60 Exposure pathways include external exposure, inhalation, and ingestion. The 95th percentile value of the simulations results for each radionuclide was chosen to calculate the DCGLs.

6.2.3 Shipbreaking Exposure Scenarios The licensee designated fourteen (14) exposure scenarios related to the shipbreaking end-state condition as significant. Two work exposure scenarios, the Remediation of hazardous materials on ship, herein referred to as Remediation Worker, and Component removal/metal cutting on ship, herein referred to as Component Removal Worker were developed and modeled by the licensee. Both exposure scenarios assume an exposure time of 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> per year (six hours per day for 250 days per year). The value for the 95th percentile of the simulation results was chosen to calculate the surface contamination limits. Detailed input parameters for external exposure, ingestion, and inhalation pathways for both models are included in Sections 6.9.2 and 6.9.3 of the LTP.

Section 6.6, Exposure Scenarios Evaluated, of the LTP describes that in the less likely but plausible scenario that the NS Savannah is reefed after license termination, the immediate shipbreaking scenario is considered a bounding scenario given the likely need to remove the structures and components containing residual radioactivity as part of the ship preparations for reefing.

As discussed in 6.9.4, Scrap Steel Scenarios in NUREG-1640, the other twelve (12) exposure scenarios, defined in NUREG-1640 Volume 1 and Volume 2, relate to steel handling and processing and groundwater infiltration by leachate from landfills or storage piles. The seven (7) steel handling and processing scenarios, herein referred to as Scrap Yard/Foundry Worker scenarios, were developed and published by the NRC staff in NUREG-1640. The exposure scenarios represent worker activities in handling and processing the steel, slag, and electric arc furnace dust. For a scrap yard worker, the exposure scenarios assume a range of external and internal exposure times from four to six hours per day for 250 days per year. For a foundry/slag worker, exposure scenarios assume a range of external and internal exposure times from two to six hours per day for 250 days per year. The value for the 95th percentile of the pathways results was chosen to calculate the surface contamination limits.

One of the Scrap Yard/Foundry Worker scenarios, titled the Scrap Yard worker, depicts the processing of steel scrap at the scrap yard (e.g., steel recycling facility). More specifically, the receptor, a scrap yard worker, may be exposed to residual radioactivity while shearing or torch-cutting the metal, briquetting or crushing thin and lightweight materials, and baling. The scrap yard scenario begins anywhere from three to 17 days after the scrap has been cleared, and steel scrap is typically shipped out within one week of arrival at the scrap yard. As such, the exposure duration is limited to the time that a worker would reasonably spend working with or in close proximity to the cleared scrap while it is at the scrap yard. The scrap yard worker is assumed to be an average of two meters from the hemispheric pile of scrap, and a uniform distribution from 0.5 to 1 is included to represent special variability of the worker.

NUREG-1640 also provides analyses related to groundwater infiltration by leachate from piles of steel scrap yards and/or recycling facilities, herein referred to as Leachate scenarios. These exposure scenarios assume an exposure time between two to six hours per day for 250 days per year. The maximum 95th percentile Effective Dose Equivalent (EDE) for each radionuclide was converted to (/) (/2). Section 6.9.4 of the LTP contains summary information of the exposure scenarios evaluated in NUREG-1640.

61 One of the Leachate scenarios, titled the Leachate industrial scrap scenario, depicts the exposures of individuals that obtain their drinking water from wells that are down gradient from industrial landfills used to dispose of cleared scrap. The only pathway addressed by these scenarios is the consumption of drinking water from the contaminated well; all other pathways have been eliminated.

Table 3.22, Normalized surficial effective dose equivalents to critical groups for steel in NUREG-1640 presents the 5th, 50th, and 95th percentile EDE of the most limiting exposure scenario for each radionuclide identified in units of

/

2. Table 6-16, Scenario Results, in the LTP converts these values in units of

/

2 by dividing by a factor of 600 (10

x 60

).

Table 6-2 below presents the bounding exposure scenario and associated dose rate coefficient for each radionuclide analyzed. The Component Removal Worker scenario was bounding for five radionuclides, while the Scrap Yard/Foundry Worker scenario and the leachate scenario are bounding for three radionuclides each. As mentioned in Section 5.2 of this SER, the three ROCs are Co-60, Cs-137, and Ni-63. Table 6-2 has been adapted from Tables 6-14 through 6-16 in the LTP.

Table 6-2 Exposure Scenario Results (/) (/)

Radionu clide Tour Guide Comp.

Rem-oval Worker Remedi ation Worker NUREG-1640 Scrap Yard/

Foundry Worker NUREG-1640 Leachate Maximum Scenario Ag-108m 9.56E-09 2.23E-02 2.21E-02 3.50E-02 0.00E+0 3.50E-02 NUREG-1640 Scrap Yard/Foundry Worker (Scrap Yard)

C-14 1.96E-09 7.08E-06 6.29E-06 1.50E-06 1.58E-05 1.58E-05 NUREG-1640 Leachate (Industrial scrap)

Co-60 1.88E-09 3.07E-02 3.06E-02 6.33E-02 0.00E+0 6.33E-02 NUREG-1640 Scrap Yard/Foundry Worker (Scrap Yard)

Cs-137 4.04E-09 7.81E-03 7.80E-03 1.25E-02 0.00E+0 1.25E-02 NUREG-1640 Scrap Yard/Foundry Worker (Scrap Yard)

Ni-63 9.99E-11 5.93E-06 1.79E-06 5.83E-07 0.00E+0 5.93E-06 Comp Removal Worker

62 Sr-90 7.10E-09 1.26E-03 4.39E-04 3.33E-04 4.83E-05 1.26E-03 Comp Removal Worker Tc-99 2.37E-09 9.13E-06 4.46E-06 2.00E-06 2.33E-03 2.33E-03 NUREG-1640 Leachate (Industrial scrap)

H-3 4.53E-12 2.17E-07 1.93E-07 4.50E-08 2.33E-05 2.33E-05 NUREG-1640 Leachate (Industrial scrap)

Fe-55 8.21E-12 3.26E-06 1.85E-06 5.67E-07 2.50E-28 3.26E-06 Comp Removal Worker Am-241 4.80E-07 3.58E-01 1.52E-02 4.83E-02 0.00E+0 3.58E-01 Comp Removal Worker Pu-239/240 9.31E-07 3.46E-01 1.47E-02 3.50E-02 0.00E+0 3.46E-01 Comp Removal Worker 6.2.4 NRC Evaluation of Exposure Scenarios, Critical Group, and Pathways The NRC staff finds that the licensee provided an adequate basis for using the Component Removal Worker, the NUREG-1640 Scrap Yard/Foundry Worker, and the NUREG-1640 Leachate scenarios to demonstrate compliance with the unrestricted release criteria of 10 CFR 20.1402, for ship surfaces and offsite soil and groundwater. These building and land use scenarios cover all potential exposure pathways, including offsite analysis, which is not reflective of site conditions at license termination but rather may occur in the future. Therefore, the conservative assumptions and scenarios in the analysis produced more restrictive DCGLs than may be expected at license termination.

The definition of critical group in 10 CFR 20.1003 is the group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances. The NRC staff concludes that the licensees use of the Component Removal Worker, the NUREG-1640 Scrap Yard/Foundry Worker, and the NUREG-1640 Leachate scenarios as reasonably foreseeable exposure scenarios and associated critical groups is acceptable.

In the preservation end-state condition, no immediate work must be done to dismantle the ship.

Therefore, receptors in the associated exposure scenarios benefit from radiological decay over time. The Tour Guide Scenario models employees that remain onboard the ship to provide walking tours, resulting in much lower exposure than other scenarios due to increased distance from source(s) of residual radioactivity.

The NRC staff agrees that immediate shipbreaking and scrapping of the NS Savannah following license termination results in the highest dose, specifically for workers. In the Remediation and Component Removal scenarios, workers would begin dismantling the ship at the time of license termination to begin removing hazardous materials (e.g., asbestos) and existing components and piping. As such, these scenarios do not take credit for any time delay and decay of residual radioactivity. The receptor would be closer to residual radioactivity during active dismantlement of the ship, resulting in greater external exposure. The NRC staff has concluded that the conceptual models, including pathways and parameters chosen, for the Remediation and Component Removal scenarios are acceptable.

63 The NRC staff recognizes that the licensee has included offsite analyses from NUREG-1640 in its evaluation of exposure scenarios to encompass multiple reasonably foreseeable end-state conditions of the NS Savannah. At license termination, the licensee expects that the ship will remain intact. Nonetheless, the licensee has included NRC published evaluations, presented in NUREG-1640, of twelve additional exposure scenarios to further evaluate the shipbreaking end-state condition in addition to its Remediation and Component Removal scenarios.

These NRC published scenarios were used to further analyze how an average member of these additional critical groups could potentially interact with the residual radioactivity left on the ship at license termination in the event that the ship is scrapped. NUREG-1757 Volume 2, Revision 2, indicates that licensees may use the generic offsite analysis presented in NUREG-1640 to screen the importance of offsite uses.

Moreover, it indicates that licensees may be able to use information from NUREG-1640 to screen their potential exposure scenarios with quantitative and/or qualitative arguments to demonstrate that the dose from certain exposure scenarios is bounded by the dose of higher-level exposure scenarios. As presented by the licensee, the dose from the exposure scenarios presented in NUREG-1640 is bounding for ROCs such as Cs-137 and Co-60.

However, the licensee is not using the exposure scenarios in NUREG-1640 as screening for potential exposure scenarios but rather as direct inputs to their DCGLs. Moreover, the licensee did not evaluate changes in recycling and disposal methods since NUREG-1640 was originally published in 2003. Therefore, it is risk-significant to confirm that the assumptions in NUREG-1640 remain valid, realistic, conservative, and/or generally applicable more than twenty years after the analysis was published.

Three ROCs have been identified for the FSS (Cs-137, Co-60, and Ni-63). Because the DCGLs for these ROCs were limited by either the Scrap Yard scenario or the licensee developed Component Removal Worker scenario, the NRC staff has performed confirmatory analysis of the Scrap Yard scenario presented in NUREG-1640. The NRC staff has compared the following parameters to modern day steel scrapping and recycling techniques and amounts as they relate to the NS Savannah.

Assumed rates of dilution in the environment Assumed rates of mixing of NRC-licensed facility material and virgin metal Percent of ferrous material recycled Recycle rate Recycling and handling methods, including distance from scrap pile Recycling activities Amount of material scrapped Although there have been changes to the steel scrapping and recycling industry and potential behavioral changes in potential workers, these discrepancies are not significantly different than the parameter values selected in the development of NUREG-1640. Additionally, the size of the ship (350 tons) is significantly smaller than the pile of scrap modeled (3500 tons). Residual contamination of the NS Savannah is limited to the Containment Vessel and Reactor Compartment but would be mixed with other ferrous material during packaging and transportation to the scrap yard facility. Then, the residual radioactivity would be further mixed with virgin metal upon arrival at the scrap yard.

64 One of the main drivers of risk in the scrap yard worker scenario is the distance to the hemispheric pile of scrap (average of two meters from the hemispheric pile of scrap, and a uniform distribution from 0.5 to 1 is included to represent special variability of the worker). The worker is assumed to be within close proximity to the scrap pile for 1000 to 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> per year, assuming that the worker also performs other tasks throughout the work day that would require them to be away from the pile. These assumptions were made based on SC&A Inc. personnel data from NRC-facilitated visits to two scrap yards processing ferrous metal scrap in 2002. As such, they are still realistic and relevant to todays scrapping practices.

The NRC staff has determined that the dose modeling and subsequent analysis of the scrap pile and associated environmental parameters are adequate to bound the DCGLs for Ni-63, Co-60 and Cs-137 for NS Savannah.

6.3 Source Term The key areas of review for the source term assumptions are the potential ROCs, configuration, residual radioactivity spatial variability, and chemical form(s) of the source. A review of the potential ROCs is included in Section 5.2 of this SER. The licensee identified four distinct source terms for the two end-state conditions of the site, which are summarized in the subsections below. The respective source term DCGL sections below describe the source term based on the relevant, risk-significant exposure scenario.

6.3.1 Tour Guide Source Term The Tour Guide exposure scenario was modeled using RESRAD-BUILD, with the assumption that the tour guide is to spend eight hours per workday, totaling 250 days per year, in the Containment Vessel and Reactor Compartment. The RESRAD-BUILD model assumes the room size to be equal to the size of the Containment Vessel: twelve meters wide by twelve meters long. Five large area circular sources with a radius of five meters each were generated. One source is on each bulkhead at the midpoint, and one is on the deck in the middle. The tour guide is assumed to be standing on the source on the deck. The dose point for the calculations is one meter above the deck. Figure 6-1, within Section 6.9, Inputs to the Scenario Calculations, of the LTP, (copied below) depicts this relationship.

65 Figure 6-1 Receptor and Sources for RESRAD-BUILD Model for Tour Guide Scenario 6.3.2 Remediation Worker and Component Removal Worker Source Term Section 6.9.2, Remediation Worker on Ship, of the LTP describes that a separate model was made for the Remediation Worker and Component Removal scenarios in MicroShield Version 8.03. As previously mentioned, both exposure scenarios assume an exposure time of six hours per workday, totaling 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> per year. A cylindrical surface was developed as the source of exposure, with a length of ten-foot pipe with the dose rate at one foot from the pipe was calculated to simulate the location of the worker. Three pipe diameters were modeled: 2 in., 4 in., and 12 in. diameter pipes. The source activity concentration was 1 Bq/cm2 for each of the radionuclides. To add conservatism, the thickness of a Schedule 40 pipe was modeled for the shields. Schedule 40 pipe has a thinner wall thickness than Schedule 80 pipe. Co-60, Ag-108m, and Cs-137 are the only gamma emitters in the list of radionuclides on the ship. The largest effective dose coefficient is from the 12 in. diameter pipe and was used in the calculations. The EDE rate, Anterior/Posterior Geometry with build-up was used to calculate the worker dose. The worker could be cutting a pipe with another pipe nearby; therefore, the EDE rate was multiplied by two for conservatism.

6.3.3 Scrap Yard/Foundry Worker and Leachate Scenarios from NUREG-1640 Source Terms Section 6.9.4 of the LTP describes that, for the Scrap Yard/Foundry Worker and Leachate scenarios presented in NUREG-1640, a hemispherical scrap pile was adopted for analysis.

Such a pile has rotational symmetry: thus, a scrap pile worker would receive the same dose regardless of their angular orientation with respect to the pile. Furthermore, a hemispherical pile is specified by one dimension, the volume, and is the simplest and most generic shape.

NUREG-1640 Appendix C states that bulk densities of ferrous scrap range from 16-22 lb/cu ft prior to compaction. Twenty pounds per cubic foot (0.32 g/cc) was adopted as the bulk density for the analysis. The source of external exposure for a scrap yard worker is a 3,500 ton pile of steel scrap, at an average distance of 2 meters from the worker. A 3,500 ton pile at a density of 20 lb/cu ft equates to a volume of 350,000 cu feet. This is equivalent to a hemisphere with a diameter at the base of approximately 100 feet.

66 Section 6.9.4, of the LTP describes that the exposed individual in the slag handling scenario is a worker at a steel mill who transfers slag using a front-end loader. The slag is spread over a large, flat area. The vehicle is either on top or at the edge of the pile. A vehicle shielding factor is applied which accounts for the shielding afforded by the loader. Based on calculated transmission factors, this parameter was assigned a triangular distribution with a range of 0.3-0.7, and a mode of 0.5. Informal calculations were performed using MicroShield and found the vehicle shielding factor range and mode to be appropriate. An uncertainty factor is applied and used to account for different locations of the worker with respect to the slag. If the worker is on top of the slag, they are exposed to an effectively infinite slab of slag, which conforms to the exposure conditions modeled in Federal Guidance Report No. 12 (ML111930454), so the uncertainty factor equals 1. When the worker is loading slag from the edge of the large, flat pile, they are exposed to one-half of an infinite slab, so the uncertainty factor equals 0.5. The uncertainty factor is assigned a uniform distribution of 0.5 to 1.

For the leachate scenarios, specifically, the source is assumed to be residually radioactive constituents of steel slag in a storage pile that leaches into the groundwater and infiltrates a drink-water well that is down gradient from the landfill. Many of the parameters of the landfill modeled in NUREG-1640 were selected from Technical Support Document: Potential Recycling of Scrap Metal from Nuclear Facilities, Part I: Radiological Assessment of Exposed Individuals published by the EPA in 2001.

6.3.4 NRC Evaluation of the Source Term Assumptions The NRC staff has evaluated the configuration, residual radioactivity spatial variability, and chemical form of the source term for each exposure scenario presented in the LTP. The NRC staff has reasonable assurance that the source terms adequately represent or are more conservative than the expected means by which an AMCG could interact with residual radioactivity from the NS Savannah.

The licensee developed Tour Guide Scenario source term is conservative. A future tour guide on the ship would not, realistically, be standing on a source in the Containment Vessel for a full workday. The NRC staff recognizes that contaminate dipping or other source material may be left within the Containment Vessel, so it is reasonable to assume that four other sources may exist in the Containment Vessel at any given time. However, the Tour Guide Scenario was not bounding for any of the radionuclides analyzed. Therefore, the NRC staff has focused its risk-informed review on the other scenarios that provided limiting DCGLs.

The licensee developed Remediation and Component Removal worker source terms are identical. The licensee includes additional conservativism, including using a thinner pipe wall thickness and an additional factor of two to represent one receptor who may be exposed to two pipes at once. The NRC staff performed confirmatory calculations to replicate the results presented by the licensee, and the results were within a small margin of error.

In its letter dated September 16, 2024, the NRC staff asked the licensee to provide any updates to the total compliance dose related to structural vs. non-structural systems on the NS Savannah. In its RAI response dated October 16, 2024, the licensee provided additional information in response RAI 5-7 on the planned end-state configuration of piping and other systems that will remain at license termination. The NRC staff has evaluated and determined that these systems that may remain at license termination have been modeled as a source term in at least one of the exposure scenarios presented.

67 Because the licensee will only have contamination of surfaces and structures, the characterization process did not consider the chemical forms of the radiological material in each of the source term materials. However, the licensee used the bounding chemical forms (i.e., the chemical form(s) that result in the highest dose per unit intake) as provided in Federal Guidance Report Number 11 (ML101590171), which is consistent with the dose modeling guidance in appendix I of NUREG-1757, Volume 2, Revision 2.

The NRC staff has reviewed the source terms presented for the various exposure scenarios in NUREG-1640 and agrees with the originally published guidance document that the scrap pile and associated environmental parameters are reasonable. For the foregoing reasons, the NRC staff finds the source term assumptions used by the licensee to be adequate.

6.4 Dose Assessment, DCGLs, and Uncertainty 6.4.1 Scrap Yard/Foundry Worker Scenario Dose Modeling Section 6.9.4 of the LTP describes that the daily duration for the scrap yard worker was assigned a range of 4-6 hours, assuming a uniform distribution, based on information obtained during visits to two scrap yards. This individual is assumed to work 250 days per year. The daily duration of external and internal exposure for the slab worker was assigned a range of 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, assuming a uniform distribution. This individual is assumed to perform other duties while being away from the slag pile. This individual is also assumed to work 250 days per year.

Inhalation rate was assigned a triangular distribution, with a mode of 1.2 m3/hr and a range of 0.6m3/hr to 1.8 m3/hr. This is based on values from the 1997 EPA Handbook and NUREG-1640.

Airborne concentration of dust that is the source of inhalation exposure for the scrap pile worker was modeled based on data collected by the U.S. Occupational Safety & Health Administration (OSHA) at the site of the World Trade Center in New York. A random sample of 250 concentrations was used to calculate the average concentration, and this parameter was assigned a custom distribution with a minimum of 0.962 mg/m3, maximum of 1.378 mg/m3. For the slag handling and processing the slag and electric arc furnace dust was a lognormal distribution with a range of 1.27 to 5.0 mg/m3 with the mode at 2.433 mg/m3. The EDE rate from all pathways for the 7 scenarios was evaluated to determine the most conservative dose rate coefficients. The 95th percentile dose rate coefficient was taken from NUREG 1640, Appendix F, Tables F1.1, F1.5, F1.9, F1.13, F1.14, and F1.22.

6.4.2 Leachate Scenario Dose Modeling There are five exposure scenarios evaluated for the leachate into groundwater in NUREG-1640.

The EDE rate from all pathways for the five scenarios was evaluated to determine the most conservative dose rate coefficients. The 95th percentile dose rate coefficient was taken from NUREG-1640, Appendix F, Tables F1.62, F1.63, F1.6, F1.65 and F1.66.

6.4.3 Component Removal Worker Scenario Dose Modeling Section 6.7, Dose Rate Equations for Remediation and Component Removal Workers, of the LTP modified equations based on those presented in NUREG-1640 to obtain dose rates, convert gm to cm2 and for varying contamination levels for the Remediation and Component Removal scenarios. A surface area factor is used for ingestion and inhalation calculations, which is the inverse of the mass to surface area. The values used in the calculations were derived from several sources, such as NUREG-1640 and ANSI N13.12-2013 and steel pipe

68 schedule charts. Values for Schedule 40 steel pipes were utilized rather than Schedule 80 pipes due to thinner walls and a more conservative assumption. A contamination factor was utilized for external exposure calculations, which accounts for varying contamination levels in pipes and components encountered during work activities. Contamination was assumed to be uniform, so the contamination factor has been conservatively set to 1. A dilution factor was utilized for inhalation calculations to account for respirator use. The equation used to calculate the dose rate from ingestion is as follows:

=

Where D is the dose rate from ingestion per unit surface activity concentration on the material for radionuclide I, F is the committed dose equivalent coefficient from figure FGR-11 for ingestion of the radionuclide I, I is the ingestion rate, f is the surface are factor, is the duration of internal exposure, is the decay constant, and is the interval from time scrap is cleared until scenario begins. Section 6.9.2 of the LTP contains the parameter selection for ingestion dose in Table 6-9, Parameters for Remediation Worker on Ship Ingestion Dose Calculations, as follows:

Table 6-9 Parameters for Remediation Worker on Ship Ingestion Dose Calculations The dose rate from inhalation equation is derived from Equation 3.8 in NUREG-1640:

=

Where is the dose rate from inhalation per unit surface activity concentration on the material for radionuclide, is the committed dose equivalent coefficient from FGR-11 for inhalation of radionuclide, is the inhalation rate, is the surface area factor, is the dilution factor used for wearing respirators, is the duration of internal exposure, is the effective dust concentration in the air, is the radioactive decay constant, is the interval from time scrap is cleared until scenario begins. Section 6.9.2 of the LTP contains the parameter selection for the

69 inhalation calculations in Table 6-10, Parameters for Remediation Worker on Ship Inhalation Dose Calculations, as follows:

Table 6-10 Parameters for Remediation Worker on Ship Inhalation Dose Calculations The only difference between the Remediation and Component Removal worker scenarios is the dilution factor. In the component removal worker scenario, the dilution factor is assumed to be one as it is assumed that the component removal workers are not wearing masks. The last equation utilized for this scenario is derived from Equation 3.6 in NUREG-1640 which is used to calculate the dose rate from external exposure:

=

Where is the dose rate from external exposure per unit surface activity concentration on the material for radionuclide i, is the EDE rate per unit surface activity concentration on the material for radionuclide i generated by MicroShield, is the exposure time, is the contamination factor, is the uncertainty factor which accounts for the variation of dose rate with position, is the radioactive decay constant, is the interval from time scrap is cleared until scenario begins. Section 6.9.2 of the LTP contains the parameter selection for the external dose calculations in Table 6-11, Parameters for Component Removal Worker on Ship External Dose Calculations, as follows:

70 Table 6-11 Parameters for Component Removal Worker on Ship External Dose Calculations Calculated surface contamination values in the remediation worker and component removal worker scenarios were developed using ModelRisk4.0. ModelRisk4.0 is a 3rd party add-on to Microsoft Excel and allows for Monte Carlo simulations of the sampling and propagation of a variety of distributions, correlations of parameter values from data sets, creation of empirical or pre-defined distributions from data sets. The number of iterations selected for executing the simulations in this analysis was set to 10,000 iterations. The 95th percentile was selected for the assessments of the remediation and component removal workers.

6.4.4 DCGLs for Radionuclides of Concern Section 6.10 of the LTP provides the results of the surface DCGLs corresponding to the maximum effective dose equivalent rate coefficient and the surface DCGLs corresponding to 15 mrem/yr that the licensee is proposing for the NS Savannah. The NRC staff edited Table 6-17 of the LTP to only display the surface contamination limits of the ROCs as found in Table 6-3, Structural Surface DCGLs.

Table 6-3 Surface Contamination Limits (DCGLs)

Radionuclide Maximum (mrem/yr)/(dpm/cm) 15 mrem/yr limit (dpm/100cm2)

Co-60 6.33E-02 2.37E+04 Cs-137 1.25E-02 1.20E+05 Ni-63 5.93E-06 2.53E+08 6.4.5 NRC Evaluation and Independent Analysis of DCGLs and Uncertainty

71 The NRC staff reviewed the scenario and parameters the licensee used to develop the DCGLs for ship structures left at license termination and finds that the use of the maximum dose from any of the scenarios evaluated is an acceptable approach for establishing DCGLs because it provides additional conservatism, regardless of the end-state configuration of the site. The NRC staff finds that the parameter values used for modeling the physical configuration of the source term are acceptable because they are either consistent with the physical configuration of the ship or scrapyard and are conservative. The NRC staff performed independent calculations of the DCGLs for the limiting scenarios, namely the Tour Guide and Component Removal Worker scenarios, and obtained comparable results as the licensee. For these reasons, the NRC staff concludes that the DCGL values proposed by the licensee for the surfaces of the NS Savannah are acceptable for demonstrating compliance with the unrestricted release criteria in 10 CFR 20.1402. Section 6.10 of the LTP provides the results of the surface DCGLs corresponding to the maximum effective dose equivalent rate coefficient and the surface DCGLs corresponding to 15 mrem/yr that the licensee is proposing for the NS Savannah. The NRC staff edited Table 6-17 of the LTP to display the 25 mrem/yr dose limit. The results are found in Table 6-4, DCGLs for all radionuclides.

Table 6-4 DCGLs for all radionuclides Radionuclide Maximum (mrem/y)/(dpm/cm) 15 mrem/yr adopted dose standard (dpm/100cm2) 25 mrem/yr dose limit (dpm/100cm2)

Ag-108m 3.50E-02 4.29E+04 7.14E+04 C-14 1.58E-05 9.47E+07 1.58E+08 Co-60 6.33E-02 2.37E+04 3.95E+04 Cs-137 1.25E-02 1.20E+05 2.00E+05 Ni-63 5.93E-06 2.53E+08 4.22E+08 Sr-90 1.26E-03 1.19E+06 1.98E+06 Tc-99 2.33E-03 6.43E+05 1.07E+06 H-3 2.33E-05 6.43E+07 1.07E+08 Fe-55 3.26E-06 4.60E+08 7.67E+08 Am-241 3.58E-01 4.19E+03 6.98E+03 Pu-239/240 3.46E-01 4.34E+03 7.23E+03 Because two of the three ROCs have DCGLs based on exposure scenarios established in existing NRC guidance, the NRC staff did not request additional uncertainty analyses to be performed on the Scrap Yard/Foundry Worker or Leachate exposure scenarios. While some of the parameters (e.g., variability of dose rate with position of the receptor to the source) in the evaluation were assigned an uncertainty factor, uncertainties due to variability in the size, density, or chemical composition of the radiation source are not addressed in NUREG-1640.

These analyses are not site-specific and therefore are not driven by a limited number of physical parameters selected by the licensee. As described in Section 6.2.4 of this SER, there are

72 additional conservativisms in parameter selection presented in NUREG-1640 that provide an adequate safety margin, regardless of changes in behavioral parameters or industry practices since the publication of the evaluations.

Moreover, the added safety margin of the licensee-adopted 15 mrem/yr dose standard gives the NRC staff reasonable assurance that any uncertainties in the dose models (licensee-and NRC-developed) are adequately represented in the DCGLs.

6.5 Alternate Scenarios: Less Likely but Plausible Section 6.6, of the LTP analyzed two less likely but plausible exposure scenarios related to artificial reefing of the NS Savannah after license termination. The National Oceanographic and Atmospheric Administration National Fishing Act of 1984 describes artificial reefing as a structure which is constructed or placed in waters covered under this title for the purpose of enhancing fishery resources and commercial and recreational fishing opportunities.

6.5.1 Diver Exposure Scenario Section 6.11.1, Diver on a Reefed Ship, of the LTP shares the licensees exposure scenario evaluations. The most likely exposed individual is a diver and the only exposure mode is through external exposure to structures and components. The same model used for external exposure to the remediation and component removal workers was used for divers with the one foot air gap changed to water. Resulting dose rates were multiplied by two for conservatism to account for impacts from two sources to the same receptor. The exposure time was conservatively assumed to be 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> a year. Table 6-18, Reefing Diver Dose Rate Coefficients and DCGLs, of the LTP presents the calculated dose rate coefficients and 15 mrem/yr activity concentrations. The NRC staff adapted Table 6-18 to shows the dose rate coefficients and associated activity concentrations. This information is shown in Table 6-5, Less Likely But Plausible Diver Exposure Scenario DCGLs.

Table 6-5: Less Likely But Plausible Diver Exposure Scenario DCGLs Radionuclide 15 mrem/yr adopted dose standard (dpm/100cm2) 25 mrem/yr dose limit (dpm/100cm2)

Ag-108m 6.85E+05 1.14E+06 Co-60 3.14E+05 5.23E+05 Cs-137 1.89E+06 3.15E+06 6.5.2 Contaminated Seafood Exposure Scenario Section 6.11.2, Consuming Fish from a Reefed Ship, of the LTP describes that the second exposure pathway evaluated in the reefing scenario is through the consumption of contaminated seafood. Current nuclide bio-accumulation models require site-specific measurements of the biomasses of various functional groups and key species. The undetermined reefing location and resulting biota composition make such modeling impossible. Using conservative assumptions of the composition of residual nuclides at the time of reefing, it can be shown that concentrations of the residual nuclides in the environment would be less than the NRC unrestricted effluent release limits. By demonstrating concentrations lower than the unrestricted effluent release limits, it is also demonstrated that concentrations of the nuclides would not create deleterious environmental effects including bio-accumulation of nuclides in species eaten by humans.

Ultimately this demonstrates that an exposure pathway via contaminated seafood is not of concern. The following conservative assumptions were made:

73 1.

The total residual man-made radioactive material is equal to one curie.

2.

All contamination is located within the containment vessel.

3.

All contamination is located within 1 mm or less of the exposure surface of the containment vessels structural steel.

4.

Structural steel deterioration is dependent on multiple factors. Two primary factors are water temperature and depth of the steel below the water surface.

a.

Higher water temperatures are correlated with faster rates of corrosion. This scenario will assume the vessel in reefed in U.S. controlled tropical waters and thus subject to higher rates of corrosion. Corrosion rates are derived from measurements collected from the USS Arizona in Pearl Harbor, Hawaii.

b.

Lower depth is associated with slower rates of corrosion. The reefing scenario will assume the highest point of the reefed vessel is 30 feet below sea level.

Correspondingly, the reactor hatch will be at depth of 56 feet and the lowest portion of the reactor compartment at a depth of 110 feet. Corrosion rates for a depth of 56 feet, the highest portion of the containment vessel, will be used for these calculations.

c.

The corrosion rate (C) of 0.0026 mm/yr will be used as calculated from the following formula, where i is the corrosion rate in mils/yr

= 0.0254/

= 0.051() + 2.96 5.

The volume of the containment vessel is estimated at 1.13E9 mL.

6.

A 100% exchange of water within the Containment Vessel is achieved at least once every five years. This is the most conservative of assumptions. All openings on the Savanah would be opened or removed prior to reefing. In such scenario, water exchange in the Containment Vessel would likely be multiple times daily.

Using the assumption of one curie of residual material and the fractions of the total activity of the nuclides, the total remaining microcuries of each nuclide can be calculated. Based on the assumption of 0.0026 mm/yr deterioration of steel in seawater and the assumption of 1mm depth of contamination, the release rate each nuclide every five years is calculated to be 0.0013 mm. This is 1.3% of contaminated steel. Based on this rate of deterioration, it is shown the concentration of the nuclides after five years of sustained release with no water exchange. This is calculated as:

=

0.013 Where V is the containment vessel volume. The activity concentrations were compared to 10 CFR 20 Appendix B, Table 2, Column 2 Effluent Concentrations. The results of the concentrations are shown in Table 6-6 below, presented as Table 6-19 of the LTP.

Table 6-6: Concentrations of the Residual Nuclides in the Submerged Containment Volume Compared with NRC Effluent Release Limits in 10 CFR Part 20 Appendix B

74 6.5.3 NRC Evaluation of Alternate Scenarios: Less Likely but Plausible According to NUREG 1757, Volume 2, Revision 2, less likely but plausible exposure scenarios are those that are possible, based on historical uses or trends, but are not likely within the next 100 years, considering current area land use plans and trends. The evaluation of less likely but plausible exposure scenarios ensures that, if land uses other than the reasonably foreseeable land use were to occur in the future, unacceptably high risks would not result.

This scenario is less likely but plausible because it is not reasonably foreseeable that the NS Savannah would be used for artificial reefing. Of the 231 ships the licensee has disposed of between 2001-2020, only four have been reefed. Because the technical approach to reefing is similar to that of shipbreaking, this end-state condition has been considered less likely but plausible. It is not unlikely that the end-state condition of the NS Savannah would be an artificial reef, and the NRC staff agrees that artificial reefing would be less likely than the others presented for detailed analysis but could be plausible over the lifetime of the ship.

For the two less likely but plausible scenarios related to the artificial reefing end-state condition, the licensee chose critical groups of a diver receiving external exposure from residual contamination on the reefed ship and a receptor consuming contaminated seafood. The diver scenario would be the most likely method of interaction with residual radioactivity. The receptor is modeled to be inside the submerged Containment Vessel swimming 1 foot away from the source for 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> per year (more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per day), which are two conservative assumptions. The DCGLs for this scenario are significantly (14-16x) greater than those for the shipbreaking scenarios, making the shipbreaking DCGLs bounding.

The contaminated seafood scenario evaluated potential impacts to the public of the deteriorating Containment Vessel, potentially contaminating the water around the artificial reef.

The licensee conservatively assumed that the water inside the Containment Vessel exchanges once every five years, such that the radionuclides released from the Containment Vessel would accumulate inside the Containment Vessel for such time. The results of this calculation yielded concentrations least one order of magnitude below the effluent release concentrations in

75 10 CFR Part 20 Appendix B for each radionuclide identified. As a result, this scenario does not take additional credit for the fraction of uptake into seafood and subsequent fraction of uptake into the receptor from the contaminated seafood. As such, the NRC staff concludes that the licensee has demonstrated that the less likely end-state condition of artificial reefing would not cause unacceptably high risks.

6.6 NRC Evaluation of the Compliance with Radiological Criteria for License Termination Conclusions The NRC staff evaluated the information provided in the LTP pertaining to the licensees assessment of the potential doses resulting from exposure to residual radioactivity remaining at the end of the decommissioning process. This evaluation was conducted using Section 2.6 of NUREG-1700, which refers to Section 5.2 and Appendix I of NUREG-1757, Volume 2, Revision 2, to conclude that the LTP is in compliance with the unrestricted release criteria specified in 10 CFR 20.1402.

The NRC staff has reasonable assurance of the following:

The licensee has adequately characterized and applied its source term.

The licensee has analyzed the appropriate end-state conditions and associated dose scenarios.

The exposure groups adequately represent a critical group.

The mathematical method and parameters used are appropriate for the dose scenarios, and parameter uncertainty has been adequately addressed.

The highest dose to the AMCG for any of the evaluated exposure scenario(s) was used to calculate the DCGLs.

The licensee has committed to using radionuclide specific DCGLs and will ensure that the total dose from all radionuclides and all sources will meet the requirements of Subpart E of 10 CFR Part 20 using the SOF approach and the compliance dose equation, =

6-4, in Section 6.1 of this SER.

The licensee has adopted an administrative dose standard of 15 mrem/yr TEDE. The DCGLs presented in the LTP are sufficient to meet the 25 mrem/yr TEDE dose limit for unrestricted release.

The NRC staff agrees that proposed DCGLs will facilitate compliance with the 25 mrem/yr TEDE plus ALARA criteria in 10 CFR 20.1402. The NRC staff has reasonable assurance that this ALARA practice is sufficient to account for any discrepancies in parameter selection, sensitivity analysis, or changes in steel scrapping/recycling since the publishing of NUREG-1640.

7 SITE-SPECIFIC COST ESTIMATE In accordance with the requirements of 10 CFR 50.82(a)(9)(ii)(F) and consistent with the guidance of RG 1.179, Chapter 7, Update of the Site-Specific Decommissioning Costs, of the NS Savannah LTP provides an updated site-specific estimate of remaining decommissioning costs to complete the D&D activities at the NS Savannah site. This portion of the LTP estimates the decommissioning costs remaining at the time of LTP submittal and compares the estimated remaining costs with the present funds set aside for decommissioning. If there is a deficit in

76 present funding, then the LTP will indicate the means for ensuring adequate funds to complete the decommissioning activities.

7.1 Financial Requirements and Cost Estimate Criteria 10 CFR 50.82(a)(9) states in part: All power reactor licensees must submit an application for termination of license. The application for termination of license must be accompanied or preceded by a license termination plan to be submitted for NRC approval.

According to 10 CFR 50.82(a)(9)(ii)(F), for NRC to evaluate and provide approval of the LTP, the submittal should include an updated site-specific estimate of remaining decommissioning costs.

In accordance with 10 CFR 50.82(a)(9)(ii)(F), Chapter 7 of the LTP provides an updated estimate of the remaining decommissioning costs for releasing the NS Savannah site for unrestricted use. This chapter also compares the estimated remaining cost with the funds currently available in the decommissioning trust fund.

7.2 NRC Evaluation of Evaluation of the Updated Site-Specific Decommissioning Cost Estimate As required by 10 CFR 50.82(a)(9)(ii)(F), the licensee estimated the remaining decommissioning costs associated with the termination of the NS Savannah license to be

$119.933 million as of December 30, 2023. The licensee is a modal agency of the United States Department of Transportation (DOT). It is a Federal licensee as defined by the NRC. As such, funds for decommissioning and termination of the NS Savannah license are provided to the licensee by Federal appropriations. This estimate is derived from Table 7-1, Estimated Remaining Decommissioning Costs as of September 30, 2023, in the LTP. The unobligated balance of funds (i.e., carryover) of $18.5 million is deducted from the appropriation of

$131 million. The result is $112.483 million, to which the remaining DECON cost of $7.45 million, as projected in Table 7-1 in the LTP, is added.

In requesting funds to support the NS Savannah, the licensee distinguished baseline (protective storage) activities from dismantlement and license termination activities, and each received its own line of accounting within the licensee's Ship Disposal Program appropriation. As required by 10 CFR 50.82(a)(8)(v) and 10 CFR 50.82(a)(8)(vii), the licensee must submit an annual report to the NRC on the status of decommissioning expenditures, remaining costs, and funding assurance levels, as well as the status of funding for managing irradiated fuel. The licensee has submitted the necessary annual reports on decommissioning cost estimates to comply with the regulations. Funds for baseline activities are requested and appropriated annually, and the funds are non-expiring. Funds for dismantlement and license termination are referred to as DECON-LT following the customary NRC definition of these terms.

The DECON-LT funds were appropriated in Fiscal Year (FY) 2017 and FY2018 as described in the Annual Decommissioning Funds Status Report for Calendar Year (CY) 2018 (ML19091A087). In March 2018, U.S. Congress provided $107 million, in addition to $24 million appropriated in FY2017 for a total of $131 million to complete decommissioning and terminate the NS Savannah license. The DECON-LT funds are also available until expended. Although there are two lines of accounting that distinguish DECON-LT from baseline activities, the licensee considers both lines to be sources of decommissioning funding and reports accordingly.

77 Full-year appropriations for FY2024 were enacted on May 8, 2024. The appropriation includes

$3 million for annual protective storage funding. The appropriation also permanently rescinds

$3.66 million of prior year unobligated balances in the Ship Disposal Program appropriations.

The final allocation of that rescission between the licensee's Ship Disposal and NS Savannah DECON accounts has not been determined as of the date of April 1, 2024, in the 2023 Annual Decommissioning Funds Status Report (ML24094A050). This report conservatively assumes that the rescission is taken wholly from the approximately $18.5 million in unobligated DECON funds available as of March 8, 2024. The protective storage account unobligated balance of approximately $0.4 million will not be applied to the rescission. The President's FY2025 Budget Request was submitted to the Congress on March 11, 2024. It includes protective storage funding at $3 million. The appropriations for FY2025 will be included in the CY2024 report.

Section 7.2.2.2, Cost Elements of the Decommissioning Estimate in relation to the Fixed Price Services Contract, of the LTP provides that there are seven cost elements of the decommissioning estimated in relation to the fixed price services contract. The following elements are not meant to be all-inclusive:

(1) Cost assumptions used, including a contingency factor; (2) Major decommissioning activities and tasks; (3) Unit cost factors; (4) Estimated costs of decontamination and removal of equipment and structures; (5) Estimated costs of waste disposal, including applicable disposal site surcharges; (6) Estimated final survey costs; and, (7) Estimated total costs.

The licensees decommissioning cost estimate was provided in Revision 1 to its PSDAR. That decommissioning cost estimate was prepared for budget planning purposes, and not as an engineering estimate. To-date, there have been no substantive changes to the previously docketed decommissioning cost estimate, other than routine escalation over time, and reporting of actual costs.

7.3 NRC Evaluation of Evaluation of the Decommissioning Funding Plan In addition to the requirement to provide an updated site-specific estimate of remaining decommissioning costs for the evaluation of the LTP, the licensee is also required to provide annual decommissioning funding status reports for the NS Savannah. Pursuant to 10 CFR 50.82(a)(8)(v), 10 CFR 50.82(a)(8)(vii), and 10 CFR 50.75(f)(1), the licensee is required to provide specific information related to decommissioning costs, expenditures, and funds.

The fixed price nature of the projected costs for decommissioning the NS Savannah were determined in CY2021 when the fixed price primary decommissioning services contract was awarded. This contract is in the same general form and scope as previous Technical Support for Integrated Management of Licensed Activities contracts awarded by the licensee and known by the acronym TSIM. Where it appears in the LTP, TSIM refers to the 2021 primary decommissioning services contract.

Section 7 of the LTP provides that the TSIM contract is such that any estimate of remaining costs on any given date is simply the value of future obligations or future invoices as of that date. Consequently, the licensee believes that new estimates are not required to meet the underlying intent of the regulation and has therefore not prepared a new engineering cost

78 estimate. The LTP summarizes the estimated remaining, unobligated costs associated with the TSIM and other service contracts that support decommissioning.

The licensee is presently engaged in the final phases of its decommissioning process and anticipates terminating its license no earlier than December 2025. The NRC staff calculated the following amounts based on the information provided in Chapter 7 of the LTP and the licensees most recent Annual Decommissioning Funds Status Report. The firm fixed price DECON services contract sets a present limit on decommissioning costs out to license termination. The amount of decommissioning funds estimated to be required for current licensed activities is approximately $3 million annually. As of September 30, 2023, the DECON-LT appropriation balance of funds available is $18.5 million. As of September 30, 2023, the licensees annual baseline activities (protective storage) appropriation balance of funds available is $700,000 for a total of $19.2 million available. The total expenditure for decommissioning and de-radiation activities is 17.1 million dollars. That leads with a positive balance available appropriation balance of about $8 million.

As a Federal licensee, 10 CFR 50. 75(e)(l)(iv) allows funding for MARADs decommissioning activities to be obtained by appropriations when necessary. As previously described, the licensee has received full funding to support its DECON-LT activities and continues to receive annual appropriations to support its baseline activities. These funds are non-expiring. The licensee currently anticipates continuing to request protective storage funding through FY2027.

For both the DECON and protective storage accounts, funds will carryover and will be available until expended.

There are two recurring commitments regarding decommissioning funding that are reported on in each Annual Decommissioning Funds Status Report. The annual submittal of new estimate commitment was established in the licensee's 2010 Decommissioning Funds Status Report (ML110940076). It was slightly revised in 2011, and reads as follows:

Submit a new estimate annually by either revising the site-specific estimate based on circumstances that affect its underlying assumptions, or by using cost escalation factors no smaller than those in the most recent revision to NUREG 1307.

This commitment remains in effect. The licensee has noted in the 2023 Decommissioning Funds Status Report an error in the reporting of this commitment beginning with the 2021 Decommissioning Funds Status Report (ML22091A289), where the prior year status of the commitment was conflated with the text of the commitment. There was no explicit written commitment change, and thus the text of the commitment remains as shown above.

The 2021 report also described in detail the nature of the fixed price decommissioning service contract and stated that the cost estimate was frozen at that time pending submittal of the LTP.

Section 7.1, Introduction, of the LTP addresses this question as follows:

The fixed price nature of the TSIM contract is such that any estimate of remaining costs on any given date is simply the value of future obligations or future invoices as of that date.

Consequently, [the licensee] believes that new estimates are not required to meet the underlying intent of the regulation...

The revised decommissioning cost estimate commitment was established in the 2010 Decommissioning Funds Status Report and read as follows:

79 The site-specific decommissioning cost estimate will be revised at least every five (5) years.

The decommissioning cost estimate was revised in 2015 and this commitment was revised in the 2019 Decommissioning Funds Status Report (ML20122A058) to read as follows:

The site-specific [decommissioning cost estimate] will be revised in the License Termination Plan.

At that time, the licensee anticipated that the LTP would be submitted in CY2021, however, due to delays caused by the COVID-19 Pandemic National Emergency, the LTP was submitted on October 23, 2023. The submittal of the LTP provides the comprehensive update of the site-specific decommissioning costs required by 10 CFR 50.82(a)(9)(ii)(F).

Per the cost analysis and evaluation in Table 7-1, Decommissioning Funds Evaluation, the NRC staff finds that adequate funds are available to decontaminate and decommission the NS Savannah.

Table 7 Decommissioning Funds Evaluation The NRC staff finds the site-specific cost estimate for remaining radiological decommissioning costs for NS Savannah is reasonable, and that the decommissioning trust fund balance, as of December 31, 2023, will be sufficient to fund the remaining radiological decommissioning expenses.

The NRC staff finds that decommissioning cost estimate and decommissioning funding plan associated with the NS Savannah are adequate and provide sufficient details associated with the funding mechanisms. After considering the evaluations presented above, it has been determined that the proposed activities can be conducted in a manner that will not pose any danger to the health and safety of the public, protect the environment, and such activities will comply with the NRC's regulations and will not negatively impact the common defense and security. The NRC staff, therefore, concludes that the licensees LTP for NS Savannah complies with 10 CFR 50.90 and 10 CFR 50.82(a)(9)(f).

2023

$19,216,532 Funds include DECON_LT appropriation balance and protective storage appropriation balance 0.00%

SOI as needed from Federal Appropriation 2026 0.00%

0.00%

Rad payment include decommissioning cost for decontamination and protectve storage projected expenses Site Restoration (2023$)

ISFSI Interim End-State (2024$)

1 0

$19,216,532

$11,150,000

$0

$0

$0 0.00%

$0

$8,066,532 2

1

$8,066,532

$3,000,000

$0

$0

$3,000,000 0.00%

$0

$8,066,532 3

2

$8,066,532

$3,000,000

$0

$0

$3,000,000 0.00%

$0

$8,066,532 4

3

$8,066,532

$3,000,000

$0

$0

$3,000,000 0.00%

$0

$8,066,532 5

4

$8,066,532

$0

$0

$0

$0 0.00%

$0

$8,066,532 6

5

$8,066,532

$0

$0

$0

$0

$0 0.00%

$0

$8,066,532 7

6

$8,066,532

$0

$0

$0

$0 0.00%

$0

$8,066,532

• Ending balance after decommissioning =

$8,066,532 Projected Earnings:

End Balance (2023$):

Beginning Balance (Oct 2023$):

Rad Payment MORAD (2023$)

Rad Payment DEF (2023$)

Federal Appropr Net Incremental Contributions &

Rate of Return on Earnings:

Nuclear Ship Savannah LTP Decommissioing/SAFSTOR Cost Analysis (Beginning Balance - Cost) * (1 + ERR) = End Balance Year /

Payment Year Reporting Year:

Current Amount Reported in DTF:

Rate of Return During SAFSTOR/Decom:

Projected End-Date of Decom Activities:

Escalation Rate (if applicable):

Real Rate of Return:

80 8

ENVIRONMENTAL CONSIDERATIONS In accordance with 10 CFR 50.82(a)(9)(ii)(G), requiring a supplement to the environmental report describing any new information or significant environmental change associated with the licensee's proposed termination activities, the licensee submitted an environmental report supplement as Chapter 8, Supplement to The Environmental Report, of the LTP. The licensee concludes that there is no new information or significant environmental changes associated with the proposed termination activities based on previous environmental analyses conducted in support of decommissioning decision-making.

In accordance with the NRCs NEPA-implementing regulations in 10 CFR Part 51, Environmental protection regulations for domestic licensing and related regulatory functions, the NRC staff prepared an EA documenting its evaluation of the radiological and non-radiological environmental impacts associated with approval of the licensees LTP. The NRC staff published a notice of the issuance of the NRCs EA and FONSI in the Federal Register on December 30, 2024 (89 FR 106612). Accordingly, the NRC staff has determined that issuance of this license amendment will not have a significant effect on the quality of the human environment.

9 PARTIAL SITE RELEASE CONSIDERATIONS No parts of the facility were released for use before approval of the LTP under 10 CFR 50.82(a)(9)(ii)(H).

10 STATE CONSULTATION In accordance with the Commissions regulations, the Maryland State official, Eva Nair, Program Manager, Air and Radiation Administration, Maryland Department of the Environment, was notified of the proposed issuance of the amendment on October 31, 2024. The State official responded by e-mail dated October 31, 2024 (ML24337A226) and did not have any comments or objections to issuing the amendment.

11 CONCLUSIONS The NRC has concluded, based on the considerations discussed above, that there is reasonable assurance that the remainder of the decommissioning activities at NS Savannah, as described in the LTP (1) will be performed in accordance with the regulations in 10 CFR Part 50; (2) will not be inimical to the common defense and security or to the health and safety of the public; and (3) will not have a significant effect on the quality of the environment.