ML25351A093
| ML25351A093 | |
| Person / Time | |
|---|---|
| Site: | NS Savannah |
| Issue date: | 12/15/2025 |
| From: | Nuclear Ship Support Services, US Dept of Transportation, Maritime Admin |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML25351A093 (0) | |
Text
U.S. Department of Transportation Maritime Administration Office of Ship Operations N.S. SAVANNAH FINAL STATUS SURVEY FINAL REPORT REPORT 1 OF 5 Revision I Approved:
Senior Technical Advisor Date:
i z./;1 /q z5.-
1 I
Prepared by:
Nuclear Ship Support Services, LLC
Final Status Survey Final Report 1 of 5, Revision 1 PREPARED BY / DATE:
PEER REVIEW / DATE:
DECOMMISSIONING PROGRAM MANAGER REVIEW / DATE:
INDEPENDENT REVIEW /
DATE:
FSS MANAGER REVIEW / DATE:
LICENSING and COMPLIANCE MANAGER REVIEW / DA TE:
SAFETY REVIEW COMMITTEE SUBMITTED/APPROVED LICENSING TERMINATION PLAN MANAGER REVIEW / DA TE:
Pete Hollenbeck, CHP John Clements, CHP Soeuth C. Soeun James Reese Daniel Mihalik John Osborne Submitted Eric Darois Rev. 1 Date Date Date Date Date Date Approved Date 2
Final Status Survey Final Report 1 of S, Revision 1 RECORD OF REVISIONS Revision Summary of Revisions Final Status Survey Final Report 1 of 5, Rev. 0 Original Corrected discrepancies identified during review of draft Report 2 of 5. The significant change is Final Status Survey Final Report 1 of 5, Rev. 1 the insertion of section 1, Introduction. The other changes are formatting corrections and table numbering. There are no change bars.
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Final Status Survey Final Report 1 of 5, Revision 1 EXECUTIVE
SUMMARY
The Nuclear Ship Savannah (NSS) Final Status Survey (FSS) program provides the necessary data and evaluation to support a finding that the ship (the licensed site) meets the Nuclear Regulatory Commission (NRC) radiological criterion for unrestricted use, thus permitting termination of the Maritime Administration's (MARAD) NRC operating license. MARAD's FSS program is described in detail in Chapters 5 and 6 of its revised License Termination Plan (LTP). In particular, Chapter 6 addresses the potential end-state condition of NSS, from among three primary disposition alternatives. These are physical destruction through shipbreaking, intentional sinking to fonn an artificial reef, and preservation for continued public use. MARAD's preferred outcome is preservation. The limiting condition for analysis is shipbreaking. In each case, the areas of the ship of greatest concern are the former Radiologically Controlled Areas (RCA) comprising the several levels of the Reactor Compartment (RC),
and the Containment Vessel (CV) which housed the major components of the nuclear steam supply system (note, several of those components were remediated and left in-situ). This FSS Final Report addresses these areas of highest concern. The results demonstrate that they meet the criterion for unrestricted use.
As described in the L TP, MARAD adopted an administrative dose limit of 15 mrem I year that is less than the NRC limit of 25 mrem/ year. Because MARAD's preferred end-state outcome is preservation, rather than immediate shipbreaking, Table 6-28 of the LTP illustrates the effect of decay over the projected seventy (70) year life of the ship post-decommissioning, beginning from 15 mrem1. At 70 years, the dose has decayed to 3 mrem As this report demonstrates, the highest residual activity dose rate inside the RC/CV was calculated to be 2.05 mrem / year. Table 6-28 is reproduced below, with the decay calculation starting at 2.05 mrem.
Table ES-1 Annual Dose Over Time (mrem)
LT 10 y 20y 30 y 40y 50 y 60 y 70 y 2.05 1.6 1.3 1.0 0.8 0.6 0.5 0.4 The NSS FSS field campaign was executed from April 2024 to January 2025. The campaign included sixty-seven (67) survey units, some of which were subdivisions of a base unit. The complete final list of survey units is provided in Table ES-4. Each survey unit has an individual Survey Unit Release Record (SURR). Each FSS Final Report, of which five (5) are planned, will include a tranche of SURRs. In this first FSS Final Report, there are fifteen (15) SURRs attached. Reports (2) through (4) will contain approximately fifteen (15) SURRs each. The fifth and last report will describe the surveys of the ship's exterior hull envelope, including the normally underwater portions of the hull. Those surveys were performed in 2019, prior to the establishment of the FSS program, when the ship was fully accessible while on drydock.
Tables ES-2 and ES-3 list the fifteen SURRs included in this report, and their residual activity dose rates.
As may be inferred from the earlier paragraphs, the maximum dose rate within the ship areas of greatest concern is 2.05 m.rem/ year, which is well below MARAD's administrative limit, and less than ten (10) percent of the NRC radiological release criterion. All fifteen survey units are acceptable for unrestricted use.
1 Both Table 6-28 and ES-1 are based on the decay of Cs-137, one of the Radioisotopes of Concern, and the one that dominates over time - see section 6.15 of the revised LTP for more detail.
Rev. 1 4
Final Status Survey Final Report 1 of 5, Revision 1 A dedicated Executive Summary may not be repeated in the remaining FSS Final Reports. Instead, the Introduction section of the follow-on reports will contain summary infonnation germane to their included SURRs.
Table ES-2 Survey Units in FSS Final Report 1 of 5 Survey Unit# MARSSIM Description of Survey Unit Class STR-101-01 I
Containment Vessel (CV) - 1st Level (Tanktop ), Starboard Side STR-101-02 1
Containment Vessel (CV) - 1st Level (Tanktop), Port Side STR-102 1
Containment Vessel (CV) - 2nd Level (Flat)
STR-103 1
Containment Vessel (CV) - 3rd Level (D Deck)
STR-104 1
Containment Vessel (CV) - 4th Level (C Deck)
STR-105-01 1
Reactor Compartment - Lower Level ( 5' - 23 ') Starboard-side Half STR-105-02 1
Reactor Compartment-Lower Level (5' - 23 ') Port-side Half STR-105-03 1
Reactor Compartment-Lower Level, Drain Wells STR-108 1
Starboard Charging Pump Room STR-109 1
Auxiliary Access Trunk, C-Deck the Cold Water Chemistry Lab (Port) and Radiation Monitoring Room (Stbd)
STR-207 2
Health Physics Lab STR-301 3
Navigation Bridge Deck - Interior Surfaces SYS-112 Impacted Primary Pressurizing System (PE) including retained portions of System the pressurizer SYS-117-01 Impacted Neutron Shield Tank Retained Walls System SYS-117-02 Impacted Fuel Transfer Tank Retained Walls System Remainder of page intentionally blank.
Rev. 1 5
Table ES-3 Final Status Survey Final Report 1 of 5, Revision 1 Residual Activity Dose Rates in Each Survey Unit Survey Unit Dose Rate Survey Unit Dose Rate (mrem/yr)
(mrem/yr)
STR-101-01 2.05 STR-108 0.15 STR-101-02 1.12 STR-109 0.10 STR-102 0.88 SYS-112 0.42 STR-103 0.39 SYS-117-01 0.37 STR-104 1.06 SYS-117-02 1.17 STR-105-01 0.12 STR-207 0.15 STR-105-02 0.14 STR-301 0.13 STR-105-03 0.27 Remainder of page intentionally blank.
Rev. 1 6
Table ES-4 SURR Designation STR-101-01 STR-101-02 STR-102 STR-103 STR-104 STR-105-01 STR-105-02 STR-105-03 STR-108 STR-109 STR-207 STR-301 SYS-112 SYS-117-01 SYS-117-02 STR-106 STR-107 STR-110 STR-111 STR-202-01 STR-205 STR-206 STR-208 STR-209 STR-211 STR-306 STR-313 STR-316 STR-317 SYS-103 Final Status Survey Final Report 1 of 5, Revision 1 Final List of NSS FSS Survey Units SURR Title Containment Vessel (CV)- 1st Level {Tanktop ), Starboard Side Containment Vessel (CV) - 1st Level (Tanktop), Port Side Containment Vessel (CV)- 2nd Level (Flat)
Containment Vessel (CV)- 3rd Level (D Deck)
Containment Vessel (CV) - 4th Level (C Deck)
Reactor Compartment - Lower Level (5' - 23 ') Starboard-side Half Reactor Compartment - Lower Level (5' - 23 ') Port-side Half Reactor Compartment-Lower Level, Drain Wells Starboard Charging Pump Room Auxiliary Access Trunk, C-Deck the Cold Water Chemistry Lab (Port) and Radiation Monitoring Room (Stbd)
Health Physics Lab Navigation Bridge Deck - Interior Surfaces Primary Pressurizing System (PE) including retained portions of the pressurizer Neutron Shield Tank Retained Walls Fuel Transfer Tank Retained Walls Port Stabilizer Room and Port Booster Pump Room Port Charging Pump Room Gas Absorption Equipment Room I Radiation Sampling Room Cupola Inner Wall, A Deck and B Deck Reactor Compartment - Dress, Undress, and Decontamination Shower Starboard Stabilizer Room Engine Room Machinery Space, Hold Through Boat Deck Horseshoe Area, Hold Deck Hot Chemistry Lab, D Deck Reactor Compartment - Mid Level, 14' Flat Deck Promenade Deck, Exterior Surf aces Refrigerator Rooms - Vegetable and Dairy (Port and Starboard Passageways and Stairwells) 14' Flat Cargo Hold No. 4 - C Deck Elevators and Elevator Shafts Soluble Poison (SP), Retained Tank Remainder of page intentionally blank.
Rev. 1 Report Number Report 1 of 5 Report 1 of 5 Report 1 of 5 Report 1 of 5 Report 1 of 5 Report 1 of 5 Report 1 of 5 Report 1 of 5 Report 1 of 5 Report 1 of 5 Report 1 of 5 Report 1 of 5 Report 1 of 5 Report 1 of 5 Report I of 5 Report 2 of 5 Report 2 of 5 Report 2 of 5 Report 2 of 5 Report 2 of 5 Report 2 of 5 Report 2 of 5 Report 2 of 5 Report 2 of 5 Report 2 of 5 Report 2 of 5 Report 2 of 5 Report2 of 5 Report 2 of 5 Report2 of 5 7
Table ES-4 STR-302 STR-303 STR-304 STR-308 STR-309 STR-311 STR-312 STR-314 STR-315 SYS-102 SYS-114-01 SYS-114-02 SYS-114-03 SYS-114-04 SYS-118 STR-201 STR-202-00 STR-203 STR-204 STR-210 STR-305 STR-307 STR-310 Final Status Survey Final Report 1 of 5, Revision 1 Final List of NSS FSS Survey Units (Continued)
Navigation Deck, Exterior Surf aces Boat Deck, Interior Surf aces Boat Deck, Exterior Surfaces A Deck Forward, Exterior Surf aces A Deck Aft, Exterior Surfaces C Deck, Interior Surfaces D Deck: Electronics Workshop, Bulle and Special Stores Rooms, Food Freezers and Refrigerators, Machinery Space Equipment and Workshop Room, Engineering Stores Room, Spare Parts Storeroom, Passageways (all)
Dry Stores (Port) and Steward (Starboard) Outboard of Horseshoe Area Cargo Hold No. 3, C Deck Emergency Cooling System (DK)
PD-TS, Contaminated Water Tank (Starboard)
PD-T6, Contaminated Water Tank (Port)
Reactor Compartment Void Tank Fresh Water Shield Tank Decommissioning Heating, Ventilation, and Air Conditioning
{HVAC)
Reactor Compartment - Mid Level, D Deck Reactor Compartment - Mid Level, C Deck Reactor Compartment - Upper Level, B Deck Reactor Compartment - Upper Level, A Deck Cargo Hold No.4-Tanktop, D Deck and CV Portal Vestibule Promenade Deck - Interior Surf aces A Deck - Interior Surf aces B Deck - Interior Surf aces SYS STR-119 Penetrations (system pipe, sleeves and structure)
SYS-106 Hydrogen Addition System (HA)
SYS-109 Intermediate Cooling System (CW)
SYS-116-01 Main Steam System including retained portions of the port side steam generator shell and associated steam drum SYS-116-02 Main Steam System including retained portions of the starboard side steam generator shell and associated steam drum Remainder of page intentionally blank.
Rev. I Report 3 of 5 Report 3 of 5 Report 3 of 5 Report 3 of 5 Report 3 of 5 Report 3 of 5 Report 3 of 5 Report 3 of 5 Report 3 of 5 Report 3 of 5 Report 3 of 5 Report 3 of 5 Report 3 of 5 Report 3 of 5 Report 3 of 5 Report4 of 5 Report4 of 5 Report 4 of 5 Report 4 of 5 Report 4 of 5 Report4 of 5 Report4 of 5 Report 4 of 5 Report 4 of 5 Report4 of 5 Report4 of 5 Report4 of 5 Report4 of 5 8
Table ES-4 FSS-310A FSS-31 IA FSS-312A FSS-313A FSS-314A FSS-315A FSS-316A FSS-317A FSS-318A Final Status Survey Final Report 1 of S, Revision 1 Final List of NSS FSS Survey Units (Continued)
External Hull, Starboard Fore External Hull, Starboard Mid External Hull, Starboard Aft External Hull, Port Fore External Hull, Port Mid External Hull, Port Aft External Hull, Starboard Above Boot Stripe External Hull, Port Above Boot Stripe External Hull, Rudder End of Section Rev. 1 Report 5 of 5 Report 5 of 5 Report 5 of 5 Report 5 of 5 Report 5 of 5 Report 5 of 5 Report 5 of 5 Report 5 of 5 Report 5 of 5 9
Final Status Survey Final Report 1 of 5, Revision 1 TABLE OF CONTENTS 1
INTRODUCTION....................................................................................................... 15 2
FINAL STA TIJS SURVEY PROGRAM OVERVIEW.................................................. 17 2.1 Survey Planning........................................................................................................... 18 2.2 Survey Design............................................................................................................. 18 2.3 Survey Implementation................................................................................................. 19 2.4 Radiological Release Criteria........................................................................................ 20 2.5 Calibration Sources and Efficiencies............................................................................. 22 2.6 Survey Data Assessment.............................................................................................. 23 2.7 Quality Assurance and Quality Control Measures........................................................... 23 2.7.1 Technician Training..................................................................................................... 24 2.7.2 Instrumentation............................................................................................................ 24 2.7.3 Periodic Surveillance.................................................................................................... 25 3
SHIP AND SURVEY UNIT IN'FORMATION.............................................................. 25 3.1 Ship Description and History........................................................................................ 25 3.2 Survey Unit Descriptions.............................................................................................. 26 3.2.1 STR-101-01 Containment Vessel (CV)- lst Level {Tank:top), Statboard Side................. 26 3.2.2 STR-101-02 Containment Vessel (CV)- lst Level (Tank:top), Port Side.......................... 26 3.2.3 STR-102 Containment Vessel (CV) - 2nd Level (Flat)................................................... 26 3.2.4 STR-103 Containment Vessel (CV)- 3rd Level (D Deck).............................................. 26 3.2.5 STR-104 Containment Vessel (CV)- 41h Level (C Deck)............................................... 27 3.2.6 STR-105-01 Reactor Compartment-Lower Level (5' -23 ') Starboard Half...................... 2 7 3.2.7 STR-105-02 Reactor Compartment-Lower Level (5 '-23 ') Port-side Half......................... 27 3.2.8 STR-105-03 Reactor Compartment-Lower Level (5 '-23 ') Drain Wells.......................... 2 7 3.2.9 STR-108 Starboard Charging Pump Room..................................................................... 27 3.2.10 STR-109 Auxiliary Access Trunk, C-Deck, Cold Water Chemistry Lab (Port) and Radiation Monitoring Room (Stbd).............................................................................................. 2 7 3.2.11 STR-207 Health Physics Lab........................................................................................ 28 3.2.12 STR-301 Navigation Bridge Deck-interior swfaces...................................................... 28 3.2.13 SYS-112 Primary Pressurizing System (PE) including retained portions of 1he pressurizer 28 3.2.14 SYS-117-01 Neutron Shield Tank wall located in 1he CV............................................... 28 3.2.15 SYS-117-02 Fuel Transfer Tank wall located in the CV.................................................. 28 3.3 Locations on General Arrangement Drawings................................................................ 29 4
SUMMARY
OF SURVEY RESULTS........................................................................ 32 4.1 Total Measurements Collected in F.ach Survey Unit....................................................... 32 4.2 Summary of Survey Results.......................................................................................... 34 4.3 Anomalous Data, Elevated Scan Results and Investigations............................................ 3 5 4.3.1 STR-101-01, Containment Vessel (CV) - 1st Level {Tanktop ), Starboard Side................. 35 4.3.2 STR-101-02, Containment Vessel (CV)- lst Level (Tanktop), Port Side......................... 35 4.3.3 STR-102, Containment Vessel (CV) - 2nd Level (Flat).................................................. 3 5 4.3.4 SYS-112, Primary Pressurizing System (PE).................................................................. 36 Rev. 1
Final Status Survey Final Report 1 of S, Revision 1 4.4 Comparison of Findings with the Release Criteria.......................................................... 36 4.5 NRC and Other Independent Verification Findings......................................................... 3 7 5
FSS FINAL REPORT SlJMMARY.............................................................................. 39 6
References................................................................................................................... 39 7
ATTACHMENTS........................................................................................................ 41 7.1 STR-101-01 Containment Vessel (CV) - 1st Level (Tanktop ), Starboatd Side.................. 41 7.2 STR-101-02 Containment Vessel (CV) - 1st Level (Tanktop ), Port Side.......................... 41 7.3 STR-102 Containment Vessel (CV) - 2nd Level (Flat).................................................... 41 7.4 STR-103 Containment Vessel (CV)-3rd Level (D Deck)............................................... 41 7.5 STR-104 Containment Vessel (CV) - 4th Level (C Deck)................................................ 41 7.6 STR-105-01 Reactor Compartment-Lower Level (5 '-23 ') Starboatd Half...................... 41 7.7 STR-105-02 Reactor Compartment-Lower Level (5 '-23 ') Port-side Half......................... 41 7.8 STR-105-03 Reactor Compartment-Lower Level (5 '-23 ') Drain Wells.......................... 41 7.9 STR-108 Starboard Oiarging Pump Room..................................................................... 41 7.10 STR-109 Auxiliary Access Trunk, C-Deck, The Cold Water Chemistry Lab (Port) and Radiation Monitoring Room (Stbd)............................................................................... 41 7.11 STR-207 Health Physics Lab........................................................................................ 41 7.12 STR-301 Navigation Bridge Deck - interior surfaces...................................................... 41 7.13 SYS-112 Primary Pressurizing System (PE) incruding retained portions of the Pressurizer 41 7.14 SYS-117-01 Neutron Shield Tank wall located in the CV............................................... 41 7.15 SYS-117-02 Fuel Transfer Tank wall located in the CV.................................................. 41 Rev. 1 11
Table 1-1 Table 2-1 Table 2-2 Table 2-3 Table 2-4 Table 4-1 Table 4-2 Table 4-3 Final Status Survey Final Report 1 of 5, Revision 1 LIST OF TABLES Survey Units in FSS Final Report 1 of 5........................................................................ 16 Traditional Scanning Covera~ Requirements................................................................ 19 Radionuclide Fractions and Nonnalized Fractions.......................................................... 20 15 mrem/y DCGI.s for ROC and Ni-63......................................................................... 21 Re-nonnalized Co-60 and Cs-137 Fractions................................................................... 22 Summary of Static Measurements................................................................................. 3 3 Summary of Random and Static Measurement Parameters.............................................. 3 4 Residual Activity Dose Rares in Each Survey Unit......................................................... 3 6 Rev. 1 12
Figure 3-1 Figure 3-2 Figure 3-3 Figure 3-4 Figure 3-5 Final Status Survey Final Report 1 of 5, Revision 1 LIST OF FIGURES NSS at Pier 13 in 2023.............................................................................................. 25 Inboard Profile......................................................................................................... 29 STR-301, Navigation Bridge Deck............................................................................ 29 STR-101-01, 101-02, 102,103,104 Containment(HistoricalDrawing)........................ 30 STR-105-01, -02 and -03, Reactor Compartment-Lower Level................................. 31 Rev. 1 13
ALARA CAD CR CV DQAP DQO DCGL ERF ETD FSS HTD ISOCS LBGR LTP MARAD MARSSIM NRC NSS ORISE QA QC RASS RC RCA RSCS ROC RPT SURR TEDE UBGR VSP Final Status Survey Final Report 1 of S, Revision 1 ABBREVIATIONS AND ACRONYMS As Low As Reasonably Achievable Computer-aided Design Contractor Report Containment Vessel Decommissioning Quality Assurance Plan Data Quality Objective Derived Concentration Guideline Level Efficiency Reduction Factor Easy-To-Detect Final Status Survey Hard-to-Detect In Situ Object Counting System Lower Bound of the Gray Region License Termination Plan United States Department of Transportation, Maritime Administration Multi-Agency Radiation Survey and Site Investigation Manual United States Nuclear Regulatory Commission Nuclear Ship SAVANNAH Oak Ridge Institute for Science and Education Quality Assurance Quality Control Remedial Action Support Survey Reactor Compartment Radiologically Controlled Areas Radiation Safety and Control Services, Inc.
Radionuclides of Concern Radiation Protection Technician Survey Unit Release Record Total Effective Dose Equivalent Upper Bound of the Gray Region Visual Sample Plan Rev. 1 14
Final Status Survey Final Report 1 of S, Revision 1 1 INTRODUCTION This Final Status Survey (FSS) Final Report is the first of five planned submittals by the Maritime Administration (MARAD) as licensee for the Nuclear Ship SAVANNAH (NSS). Each report contains a tranche of Survey Unit Release Records (SURR) as attachments. Each SURR documents the radiological status of an individual survey unit relative to the Nuclear Regulatory Commission (NRC) unrestricted use criteria, and MARAD's administrative limit. The report itself provides a summary of the survey results for the included survey units and describes principal fmdings that demonstrate that each unit is acceptable for unrestricted use.
As described in the LTP, MARAD is seeking unrestricted use oftheNSS site on the basis of the NRC 10 CFR 20.1402 radiological criteria, which defines that a site is acceptable for unrestricted use if the residual radiation that is distinguishable from background results in a Total Effective Dose Equivalent (TEDE) to an Average Member of the Critical Group that does not exceed 25 mrem per year, and that residual radioactivity has been reduced to levels that are as low as reasonably achievable (ALARA). To ensure that the NSS meets these requirements, MARAD established an administrative limit ofless than or equal to 15 mrem per year. This administrative limit includes ALARA considerations. As described herein, this administrative release criterion is translated into site-specific Derived Concentration Guideline Levels (DCGLs) for assessment and summary.
This FSS Final Report, including its attached SURRs, documents that FSS activities were performed in a manner that is consistent with the guidance provided in Revision 1 of NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) (Reference b ), as described in the L TP, and as incorporated into the NSS FSS implementing procedures listed in Section 2.1. This report is consistent with the guidance provided in NUREG-1757, Vol. 2, Consolidated Decommissioning Guidance Characterization, Survey, and Determination of Radiological Criteria -Final Report (Reference c ), and MARSSIM. This guidance is incorporated in NSS procedure STS-005-035, Preparation of Survey Unit Release Records and FSS Final Reporls (Reference d), which is the governing procedure under which these documents were prepared.
In this FSS Final Report, there are fifteen (15) Survey Units. They are listed below in Table 1-1, Survey Units in FSS Final Report 1 of 5.
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i Final Status Survey Final Report 1 of 5, Revision 1 Table 1-1 Survey Units in FSS Final Report 1 of 5 Survey Unit#
MARSSIM Description of Survey Unit Class STR-101-01 I
Containment Vessel (CV) - 1st Level (Tanktop), Starboard Side STR-101-02 I
CV - I st Level (Tanktop ), Port Side STR-102 I
CV - 2nd Level (Flat)
STR-103 I
CV - 3rd Level (D Deck)
STR-104 1
CV - 4th Level (C Deck)
STR-105-01 1
Reactor Compartment - Lower Level (5' - 23 ')
Starboard-side Half STR-105-02 I
Reactor Compartment - Lower Level (5' - 23 ')
Port-side Half STR-105-03 1
Reactor Compartment - Lower Level, Drain Wells STR-108 1
Starboard Charging Pump Room STR-109 1
Auxiliary Access Trunk, C-Deck the Cold Water Chemistry Lab (Port) and Radiation Monitoring Room (Stbd)
SYS-112 Impacted Primary Pressurizing System (PE) including retained portions of System the pressurizer SYS-117-01 Impacted Neutron Shield Tank Retained Walls System SYS-117-02 Impacted Fuel Transfer Tank Retained Walls System STR-207 2
Health Physics Lab STR-301 3
Navigation Bridge Deck - Interior Surf aces These units are described more fully in Section 3 of this report. Their relationship to the licensed site, which is defined by the exterior envelope of the ship itself, is also described in Section 3. Section 4 summarizes the survey results, and Section 5 summarizes the principal findings that demonstrate that each unit is acceptable for unrestricted use. Table 1-2, on the next page, shows the dose from residual activity in each survey unit. Each unit is well below the MARAD 15 mrem administrative limit, and the NRC release criterion of 25 mrem, and is acceptable for unrestricted use.
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Rev. I 16
r Final Status Survey Final Report 1 of 5, Revision 1 Table 1-2, Residual Activity Dose Rates in Each Survey Unit Survey Unit Dose Rate Survey Unit Dose Rate (mrem/yr)
(mrem/yr)
STR-101-01 2.05 STR-108 0.15 STR-101-02 1.12 STR-109 0.10 STR-102 0.88 SYS-112 0.42 STR-103 0.39 SYS-117-01 0.37 STR-104 1.06 SYS-117-02 1.17 STR-105-01 0.12 STR-207 0.15 STR-105-02 0.14 STR-301 0.13 STR-105-03 0.27 The dose contribution from residual activity in each survey unit was calculated by taking the mean activity from all final static measurements and dividing that value by the gross activity Derived Concentration Guideline Level to derive a mean fraction. The mean fraction was then multiplied by 15 mrem/yr to derive the dose from residual activity in the survey unit.
2 FINAL STATUS SURVEY PROGRAM OVERVIEW MARAD conducts its FSS activities in accordance with the FSS Plan described in Chapter 5 of the L TP and as described in the Decommissioning Quality Assurance Plan (DQAP) (Reference e ). A suite of FSS implementing procedures was developed (see list below), with procedure STS-005-029, Final Status Survey Program (Reference f) establishing the overarching framework for the PSS program. This procedure expands the description of the PSS organization with roles and responsibilities, and establishes requirements and methods for FSS planning, survey design and implementation, data assessment and investigation, remediation (including potential reclassification and resurvey, where and if required), and reporting. The FSS program is conducted within the overall context of the site procedure program, so that activities are subject to, among other things, the DQAP implementing procedures, Work Control Process, Health and Safety Plan, Radiation Protection Plan, etc. The FSS-specific administrative procedures are listed below:
STS-005-029, Final Status Survey Program (Reference/)
STS-005-030, Preparation of FSS Packages (Reference g)
STS-005-031, Calculation of the Number of Measurements (Reference h)
STS-005-032, Survey Unit Turnover and Control (Reference i)
STS-005-033, Final Status Survey Data Assessment (Reference j)
STS-005-034, Survey Unit Classification (Reference k)
STS-005-035, Preparation of Survey Unit Release Records (Referenced)
SIC-MA-Q-25, FSS Survey Unit Inspections After Surveys are Completed (Reference 1)
Rev. 1 17
Final Status Survey Final Report 1 of 5, Revision 1 2.1 Survey Planning The Data Quality Objectives (DQO) process is incorporated as an integral component of the data life cycle of the NSS FSS program. This process, described in MARSSIM, is a series of planning steps found to be effective in establishing criteria for data quality and developing survey plans. DQOs allow for systematic planning and are specifically designed to address problems that require a decision to be made and provide alternate actions. The DQO process is flexible (i.e., it employs a graded approach) in that the level of effort associated with planning a survey is based on the complexity of the survey and nature of the hazards. Finally, the DQO process is iterative, allowing the suivey planning team to incorporate new knowledge and modify the output of previous steps to act as input to subsequent steps.
The DQO process was used in the FSS planning phase for scoping, characterization, remediation, and FSS package development.
The DQO process consists of performing the following seven steps:
State the Problem; Identify the Decision; Identify the Inputs to the Decision; Define the Boundaries of the Decision; Develop a Decision Rule; Specify Tolerable Limits on Decision Errors; and, Optimize the Design for Obtaining Data.
These DQO process steps were completed for each FSS package.
2.2 Survey Design The survey design was a MARSSIM based approach using a combination of static measurements and scans in each survey unit. Visual Sample Plan (VSP) (Reference m) software was used to create maps and to select systematic sample locations for Class 1 and Class 2 survey units and random locations for Class 3 survey units.
The number of measurements for each survey unit was determined in accordance with STS-005-031. The relative shift for the survey unit data set is defined as shift(~), which is the Upper Boundary of the Gray Region (UBGR), or the Operational DCGLw, minus the Lower Bound of the Gray Region (LBGR),
divided by sigma ( cr), which is the standard deviation of the data set used for survey design. The optimal value for the relative shift should range between one (l) and three (3). If the relative shift is less than one (1), or greater than three (3), then the value of two (2) will be used for the relative shift.
Each Class l and Class 2 survey unit had a Remedial Action Support Survey (RASS) performed prior to turnover to the FSS staff. That data was used to both calculate sigma ( cr) and determine whether the survey unit could pass the FSS data assessment.
Class 3 survey units did not have a RASS or turnover survey performed because they inherently have a very low probability of being contaminated with licensed materials above 10% of the DCGL. Therefore, STS-005-031 specifies using a valueof0.3 as an estimate of the standard deviation. The LBGR was set at 0.5 and the DCGL set to l. The relative shift was calculated as 1.67 and rounded up to 1. 7.
The sample size in all cases was determined from Table 5.5 ofMARSSIM with a Type I and Type II enor of 0.05.
Rev. 1 18
Final Status Survey Final Report 1 of 5, Revision 1 Each survey design specified the method and location for collection of static measurements and scan coverage. Scan coverage for each survey unit followed the scanning coverage presented in LTP Table 5-3, presented as Table 2-1 below.
Table2-1 Traditional Scanning Coverage Requirements Survey Unit Classification Required Scanning Coverage Fraction Class I 100%
Class 2 Decks, or lower bulkheads of rooms: 10% to I 00%
Upper bulkheads or overheads: 10% to 50%
Class 3 Judgmental Systems and Components Judgmental Each survey design specified the collection of two (2) judgmental and (1) duplicate measurement for Quality Control (QC) purposes. Investigation Levels are also specified in the survey design.
Confirmatory measurements are required when an investigation level is exceeded.
The product of the suivey design process is an FSS package, which addresses various elements of the survey, including, but not limited to:
Maps of the survey area showing the survey unit(s) and measurement/sample locations, as appropriate; Applicable DCGLs; Instrumentation to be used; Types and quantities of measurements to be made or collected; Investigation criteria; Quality Assurance (QA) and Quality Control requirements (e.g., replicate measurements);
Applicable health and safety procedures; and, Applicable operating procedures.
2.3 Survey Implementation FSS field activities were considered work and were subject to the STS 002-011 Work Control Process procedure. Master work packages were developed, incorporating all applicable site requirements, and the work was scheduled The FSS packages described above formed the technical scope for field surveys.
Prior to performing the FSS, each survey unit was walked down and turned over to the FSS staff. This was performed under STS-005-032. Each survey unit was inspected to ensure all decommissioning activities were 1) complete and 2) any unsatisfactory conditions had been resolved Isolation and controls were implemented. Large area smear surveys in regularly accessed areas of Class 3 survey units are used to determine whether radioactive material has been reintroduced to these survey units. The controls and smears will remain in effect until license termination.
Prior to starting a survey, the survey team was briefed on the FSS package. Each package has a daily journal to record important information and logs to record static measurement and scan data. There are logs to record instrument operability checks before and after the surveys.
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Final Status Survey Final Report 1 of S, Revision 1 Radiological surveys were perf onned with the following instrumentation:
Ludlum model 3002 with a Ludlum model 43-93 dual-phosphor scintillation detector (direct static beta measurements and beta scans)
Ludlum model 3002 with a Ludlum model 43-37 gas flow proportional detector (beta scans)
Total surface activity measurements were collected with the detector in contact with the surf ace. Beta scan data with the Ludlum 43-93 detector was collected by holding the detector at approximately 0. 75 inches above the surface and moving at a speed of one (1) detector width per second. The Ludlum 43-3 7 detector was on wheels with the detector height set at 0.75 inches and moved at a speed of one (1) detector width per second.
2.4 Radiological Release Criteria The significant Radionuclides of Concern (RO Cs) identified in NSS Contractor Report (CR) -109, RC/CV Characterization Report (Reference n) were Co-60, Cs-13 7, and Ni-63. Characterization data in CR-109, measurements for alpha activity showed no significant alpha activity was present. CR-139, Calculations to SupporlNS SAVANNAH Surface Contamination DCGLs (Reference o) confirmed that Co-60, Cs-137 and Ni-63 comprise approximately 99% of the total residual radioactivity on the ship.
L TP Chapter 6, Section 6.13 discusses the process used to derive the ROC for the decommissioning of the NSS, including the elimination of insignificant dose contributors from the initial suite. Table 2-2 presents the initial suite of ROC for the decommissioning of the NSS, and the nonnalized mixture fractions based on the radionuclide mixture.
Table 2-2 Radionuclide Fractions and Normalized Fractions Nuclide Sum of Normalized Only Co, Ni Fractions Sum of and Cs Fractions C-14 3.37E-02 5.6IE-03 Co-60 9.78E-02 l.63E-02 l.63E-02 Ni-63 5.40E+o0 9.00E-01 9.00E-01 Sr-90 9.22E-06 1.54E-06 Tc-99 l.27E-04 2.1 lE-05 Ag-108m 8.30E-04 l.38E-04 Cs-137 4.57E-01 7.61E-02 7.61E-02 H-3 l.93E-03 3.21E-04 Fe-55 8.l0E-03 l.35E-03 Am-241 5.79E-07 9.66E-08 Pu-239/240 4.81E-07 8.0lE-08 Totals 6.00E+o0 l.00E+o0 9.93E-01 Table 2-3 presents the 15 mrem/y DCGLw for Co-60, Ni-63 and Cs-137 from Chapter 6 of the LTP.
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Final Status Su~vey Final Report 1 of 5, Revision 1 Table 2-3 15 mrem/y DCGLs for ROC and Ni-63 Radionuclide DCGLw (dpm/100cm2)
Co-60 2.37E+o4 Ni-63 2.53E+o8 Cs-137 1.20E-+-05 Nickel-63 is a Hard-to-Detect (HTD) radionuclide. To account for Ni-63, the Easy-To-Detect (ETD) radionuclide Co-60 will be used as a surrogate.
The surrogate DCGL is computed as:
DCGL,rrD DCGLsurrogate = DCGLETD X (fHTD£TDXDCGLETD)+DCGL,rrD Where:
Equation 1 DCGLEm DCGLmn the DCGL for the easy-to-detect radionuclide the DCGL for the hard-to-detect radionuclide FHm:EID the activity ratio of the hard-to-detect radionuclide to the easy-to-detect radionuclide Using the data in Table 2-2, theNi-63/Co-60ratio equals 55.3. Solving Equation 1 using the DCGLs for Co-60 and Ni-63 in Table 3-1 and the Ni-63/Co-60 ratio of 55.3 is shown below.
DCGLco-6o surroptc = 2.37E + 04x C 2
53 E+~a E
= 2.36E+o4 dpm/100 cm2 55.3 X 2.37E+04 +2.53 +08 A gross activity DCGLw has been established, based on the representative radionuclide mix, as follows:
Gross Activity DCGLw = ~_h_ __ _h_ Equation 2 DCGL1 DCGL2 + DCGLn Where:
f o
=
fraction of the total activity contributed by radionuclide n n
=
the number of radionuclides DCGLn = DCGL for the ROC Of the three radionuclides, only Co-60 and Cs-137 can be detected by gross activity measurements. The normalized fractions for both radionuclides must be re-normalized for use in Equation 2. Table 2-4 presents the re-normalized fractions for each radionuclide.
Remainder of page intentionally blank.
Rev. 1 21
Final Status Survey Final Report 1 of 5, Revision 1 Table 2-4 Re-normalized Co-60 and Cs-137 Fractions Nuclide Normalized Only Co-60 Re-sum of and Cs-137 normalized fractions Fractions Fractions C-14 5.61E-03 Co-60 l.63E-02 l.63E-02 l.76E-01 Ni-63 9.00E-01 Sr-90 l.54E-06 Tc-99 2.l IE-05 Ag-108m l.38E-04 Cs-137 7.61E-02 7.61E-02 8.24E-01 H-3 3.21E-04 Fe-55 l.35E-03 Am-241 9.66E-08 Pu-239/240 8.0IE-08 Totals l.00E+o0 9.24E-02 l.00E+o0 Solving Equation 2 using the re-normaliz.ed Co-60 and Cs-137 fractions and the Co-60 surrogate DCGL is shown below.
Gross Activity DCGLw = 1_76E-i ~ ~
= 6.97E+o4 dpm/100 cm2 2.368+04 1.ZOB+os 2.5 Calibration Sources and Efficiencies As defined in MARSSIM and NUREG-1507 (Reference p ), instrument efficiency is derived by measuring the surface emission rate of a clean, calibrated and certified National Institute of Standards and Technology traceable, reference source. The sources are Tc-99 (Co-60 surrogate) and Cs-13 7 with an active area of 150 cm2 (150mm by 100mm).
When addressing the issues associated with the derivation of the appropriate total efficiency for the measurement, MARSSIM states (page 6-24) that the use of a total efficiency derived from measurements made on certified 41t activity traceable sources"...is not a problem, provided that the calibration source exhibits characteristics similarto the surface contamination (i.e., radiation energy, backscatter effects, source geometry, self-absorption)." These parameters were evaluated and documented in CR-164 (Referenceq). Most of the contaminated surfaces on the ship are metal, especially in the Class 1 and Class 2 survey units. The total efficiency ( as measured by exposing the detector to the calibration sources and comparing its response to the stated total 41t activity) is appropriate for making measurements on metal surfaces, having taken into account both source efficiency and instrument efficiency.
Initial instrument beta efficiencies were determined by placing the Ludlum 43-93 detector both in contact with and at a distance of 2 cm from both reference sources. The floor monitor Ludlum 43-3 7 detector was placed at a distance of 2 cm from both the sources. The effects that various surface conditions have on detection sensitivities were evaluated by adding successive layers of paper attenuators large enough to completely cover the calibration sources. The collected data was plotted by taking the instrument Rev. 1 22
Final Status Survey Final Report 1 of 5, Revision 1 response at the attenuation thickness and dividing by the instrument response with no attenuation. An exponential trendline was then generated to define the curve for each source. The equation for the curve is identified as the Efficiency Reduction Factor (ERF).
The Historical Site Assessment, CR-003 (Referencer) describes actions taken in the 1970's to place the ship in a mothballed condition. One of the actions is listed below:
Containment vessel, port and starboard charge pump rooms, and lower reactor compartment bilges decontaminated, washed, vacuumed, and painted.
Based on this information, CR-164 recommended using a 15 mglcm2 ERF for those surfaces.
The efficiencies used to convert count rates to activity are weighted efficiencies in which the Tc-99 (Co-60 surrogate) and Cs-137 efficiencies are weighted by the fraction Co-60 and Cs-137 comprise the NSS fingerprint, as shown in Section 2.5. For example, if the 41t contact Tc-99 efficiency is l.00E-01 c/d and the 41t Cs-13 7 contact efficiency is l.87E-0 1 c/d, the weighted contact efficiency for unattenuated surfaces is shown below.
Unattenuated weighted efficiency (c/d) = (l.00E-Ol)(l.76E-0 1) + (1.87E-0 1 )(8.24E-Ol) = l.72E-01 c/d For surfaces in the Containment Vessel, port and starboard charge pump rooms, and lower reactor compartment bilges, where there could be contaminated surfaces that have been painted over, an attenuated weighted efficiency is used. Assuming the same 41t contact efficiencies as shown in the example above, the weighted contact efficiency for painted surfaces used a density thickness for paint at 15 mglcm2 is shown below and is described in CR-164.
Painted surface weighted efficiency (c/d) =
(1.00E-01)(4.00E-02) + (1.87E-01)(5.02E-01) = 9.78E-02 c/d 2.6 Survey Data Assessment Data assessment starts with a review of the collected data to ensure all plan specified data has been collected. The next step is to convert the count rate data into activity concentrations ( dpm/100 cm 2). The resulting gross activity concentrations were then compared to the gross activity DCGLw and the Investigation Levels. Any measurements greater than or equal to the Investigation Levels were either 1) investigated, remediated and re-surveyed or 2) remediated and re-surveyed. The Sign Test was then used to evaluate the total surface activity concentrations. The total surf ace activity concentrations were also displayed graphically as Quantile Plots and Histograms.
Replicate measurements were evaluated against the acceptance criteria for replicate static measurements in NUREG-1576, MARLAP Volume 1, Appendix C, Section C.4.2.2 Duplicate Analyses (Reference s).
- 2. 7 Quality Assurance and Quality Control Measures MARAD implements a comprehensive DQAP to assure conformance with established NRC requirements and accepted industry standards. All activities, including FSS activities, performed at the NSS are required to meet the requirements of the DQAP. The participants in the DQAP ensure that the FSS activities are performed in a safe and effective manner.
The DQAP makes no distinction between perfonning FSS activities and performing any other activities in the scope of the DQAP. All activities in the scope of the DQAP are performed in accordance with the DQAP. The function of FSS is an activity in the scope of the DQAP.
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Final Status Survey Final Report 1 of S, Revision 1 QC and QA measures were integrated into all decommissioning activities, including implementation of the FSS. All FSS activities essential to data quality were implemented and performed under approved procedures.
QA and QC activities for the FSS effort ensured that surveys were performed by trained individuals using approved written procedures and properly calibrated instruments that were sensitive to the suspected contaminant. In addition, QC measures were taken to obtain quantitative information to demonstrate that measurement results have the required precision and are sufficiently free of errors to accurately represent the site being investigated. QC checks were perfonned as prescribed by the implementing procedures for field measurements. For field measurements, replicate measurements in each survey unit were made for randomly chosen location(s) of the original measurements by either a different technician or by a different instrument. For Class 2 and Class 3 survey units, judgmental measurements were performed. These biased measurements were performed in locations based on the supervisor's and technician's site knowledge and professional judgment. Judgmental assessments provided added assurance that residual contamination on the ship was adequately located and characterized. The basis for judgmental assessments is documented in the FSS package for each survey unit
- 2. 7.1 Technician Training All FSS technicians employed in the MARAD FSS program were trained and qualified to perform the surveys in accordance with STS-002-001, STS Training Program, Reference (t). Training records are on file and available for review.
Over the course of the FSS program, the supply of readily qualified FSS technicians was very limited, given the high demand for such personnel across the large number of active decommissioning projects nationwide. To address this shortfall, NSS FSS technicians were sourced from the population of highly trained Radiation Protection Technicians (RPTs) that service operating plants. To prepare these individuals to perform FSSs, a specialized three-day training course for FSS technicians, identified as the FSS Academy, was developed and deployed by Radiation Safety and Control Services, Inc. (RSCS)2 in early 2024, and proof-tested on board NSS. The three instructors for this course were health physicists certified by the American Board of Health Physics, and each have over thirty years of professional health physics experience including In Situ Object Counting System (ISOCS) modeling. The three instructors are Subject Matter Experts. The two-day classroom portion was designed for managers, supervisors and technicians. It discussed an oveJView of MARSSIM, the NSS L TP and implementation of FSS packages.
On the third day of the course, prospectiveFSS technicians received specialized training on the use of instrumentation, including the !SOCS. Thetechnicians were taught how to perform a QC check as well as how to collect a measurement. The technicians only perform these two tasks. The collected spectrum, along with a picture of the measurement surf ace and the distance from the detector to the measurement surf ace are sent to a qualified subject matter expert to generate the efficiency model in the Geometry Composer software and perform the analysis. Academy RPT participants are required to pass a written exam and practical factors demonstration before qualifying as FSS Technicians.
- 2. 7.2 Instrumentation Instrumentation employed in the FSS program was calibrated, maintained, and operated in accordance with applicable NSS procedures. This included daily pre and post survey response 2 As noted in other reports to NRC, RSCS is one of two companies that form the joint venture decommissioning services contractor to MARAD. RSCS is responsible for the execution of the MARAD FSS program.
Rev. 1 24
Final Status Survey Final Report 1 of 5, Revision 1 checks, and daily pre and post use verification that instruments were in working order and within established tolerance limits.
2.7.3 Periodic Surveillance In accordance with References (i) and (l), periodic surveillance and surveys of survey units have been, and continue to be performed to verify that no new residual radioactivity has been introduced to a survey unit before the license is terminated. If contamination is discovered, STS procedures govern corrective actions.
3 SHIP AND SURVEY UNIT INFORMATION 3.1 Ship Descripdon and History The NSS was an 80 MW th pressuriz.ed-water nuclear reactor. MARAD is the owner and licensee of NSS.
MARAD Headquarters is located at 1200New Jersey Ave., SE, Washington, DC. The NSS is located at Baltimore Harbor, Pier 13 in the Canton industrial district of the port, near the Seagirt Marine Terminal in Baltimore, Maryland. The street address is4601 NewgateAvenue,Baltimore,MD 21224. See Figure 3-
- 1. The site is licensed under Possession-only License No. NS-1, Docket No. 50-23 8. The licensed site of the NSS is the boundary defined by the ship's hull.
Figure 3-1 NSS at Pier 13 in 2023 Conceived in the 1950's as part of President Eisenhower's "Atoms for Peace" program, the NSS was built by New York Shipbuilding Corporation in Camden, New Jersey. The nuclear power plant was designed and constructed by Babcock and Wilcox. The reactor was first brought to power in 1961, and the ship embarked on seagoing trials the following year. After a series of demonstration voyages, the ship carried commercial freight in an experimental program designed to demonstrate the safety and viability of nuclear-powered merchant ships.
Rev. 1 25
Final Status Survey Final Report 1 of 5, Revision 1 In 1970, the ship was removed from service, and the reactor was shut down for the final time. The reactor had been operated for approximately 21,225 effective full power hours and the ship had traveled over 450,000 miles. In 1976, the reactor and auxiliary systems were deactivated, disabled and no longer performed any active functions.
3.2 Survey Unit Descriptions This section provides a brief description of each of the survey units included in this report. Their individual release records are attached. The NSS library of original technical drawings was scanned and placed into a drawing database. These drawings were used to model the ship within Computer-aided Design (CAD) software. The CAD model of the ship was used to calculate surface areas and subsequently integrated with VSP software for use in generating the survey unit maps.
3.2.1 STR-101-01 Containment Vessel (CV)- lst Level (Tanktop), Starboard Side STR-101-0 I is an impacted Class I survey unit. The CV I st level housed the reactor plant and some supporting systems. The reactor rested on a support ring on the CV 1st level, and the entire CV rested on a cradle located on the tank top. The survey unit consists of the internal surf aces of the starboard side CV shell, ring stiffeners and the deck/new fiberglass grating to the centerline. The I st-level deck, shell, ring stiffeners, and reactor vessel foundation up to the 14-foot elevation are considered one and ref erred to as the CV shell. The I st level starboard-side half of the CV shell surface area is approximately 64.84 square meters (64.84 m2); this includes the shell surface above 2 meters. See Figures 3-2 and 3-4.
3.2.2 STR-101-02 Containment Vessel (CV)- lst Level (Tanktop), Port Side STR-101-02 is an impacted Class 1 survey unit. The port side containment vessel's 1st level housed the reactor plant and some supporting systems. The CV 1st level housed the reactor plant and some supporting systems. The reactor rested on a support ring on the CV 1st level, and the entire CV rested on a cradle located on the tank top. The survey unit consists of the internal surf aces of the port side CV shell, ring stiffeners and the deck/new fiberglass grating to the centerline. The 1st-level deck, shell, ring stiffeners, and reactor vessel foundation up to the 14-foot elevation are considered one and ref erred to as the CV shell. The 1st level port-side half of the CV shell surface area is approximately 64.84 square meters (64.84 m2); this includes the shell surface above 2 meters. See Figures 3-2 and 3-4.
3.2.3 S1R-102 Containment Vessel (CV)- 2nd Level (Flat)
STR-102 is an impacted Class 1 survey unit. The CV housed the reactor plant and some supporting systems. The 2nd level gives access to the heat exchanger man ways as well as the pressurizer heaters. The survey unit consists of the internal surfaces of the CV shell, ring stiffeners, and the deck/new fiberglass grating from the 14-foot to 23-foot elevation. The 2nd level grating surface area is approximately 40.71 square meters (40.71 m2); the total surface area, including the shell, is approximately 176.97 m2; this includes the shell surface above 2 meters. See Figures 3-2 and 3-4.
3.2.4 STR-103 Containment Vessel (CV) - 3rd Level (D Deck)
STR-103 is an impacted Class 1 survey unit. The CV housed the reactor plant and some supporting systems. The 3rd level gives access to the heat exchanger steam drums manways and the platform for the pressurizer as well as the reactor coolant pumps and motors. The survey unit consists of the internal surfaces of the CV shell, ring stiffeners, and the deck/new fiberglass grating from the 23-foot to the 32-foot elevation. The 3rd level deck surface area is approximately 62.05 square meters (62.05 m2); the total surface area, including the shell, is approximately 208.93 m2; this includes the shell surface above 2 meters. See Figures 3-2 and 3-4.
Rev. 1 26
Final Status Survey Final Report 1 of 5, Revision 1 3.2.5 STR-104 Containment Vessel (CV) - 4th Level (C Deck)
STR-104 is an impacted Class 1 survey unit. The CV housed the reactor plant and some supporting systems. The 4th level gives access to the pressurizer platform, the reactor head area, and the personnel and equipment hatches. The survey unit consists of the internal surfaces of the CV shell, ring stiffeners, and the deck/new fiberglass grating from the 32-foot to the 41-foot elevation. The 4th level deck surface area is approximately 24.57 square meters (24.57 m2); the total surface area, including the shell, is approximately 153.62 m2; this includes the shell swface above 2 meters. See Figures 3-2 and 3-4.
3.2.6 STR-105-01 Reactor Compartment - Lower Level (5 '-23 ') Starboard Half STR-105-01 is an impacted Class 1 surveyunit. The RC where the CV is located extends from 5' Elevation {Tank.top) to above the A Deck level. The RC contains support systems for the reactor plant. STR-105-01 is the Port-side to the centerline of the RC lower level, 5' - 23' elevation. The deck surface area is approximately 99.44 square meters (99.44 m2); the total surface area, including the bulkheads, raised solid deck plating and overhead, is approximately 352.16 m2 ; this includes the surfaces above 2 meters. See Figures 3-2 and 3-5.
3.2.7 STR-105-02 Reactor Compartment-Lower Level (5'-23') Port-side Half STR-105-02 is an impacted Class 1 survey unit. The RC where the Containment Vessel (CV) is located extends from 5' Elevation (Tanktop) to above the A Deck level. The RC contains support systems for the reactor plant. STR-105-02 is the Port-side to the centerline of the RC lower leveL 5' -
23' elevation. The deck surface area is approximately 99.30 square meters (99.30 m2); the total surface area, including the bulkheads, raised solid deck plating and overhead, is approximately 340.60 m2 ; this includes the surfaces above 2 meters. See Figures 3-2 and 3-5.
3.2.8 STR-105-03 Reactor Compartment - Lower Level (5'-23') Drain Wells STR-105-03 is an impacted Class 1 survey unit. There are two Drain Wells in the Reactor Compartment Lower Level, one aft and one forward. Based on CAD drawings, the Drain Wells deck surface area is approximately 2.44 square meters (2.44 m 2). The total surface area including the bulkheads and overhead of the Drain Wells is 10.5 m2. See Figures 3-2 and 3-5.
3.2.9 STR-108 Starboard Charging Pump Room STR-108 is an impacted Class 1 survey unit located on the Hold Deck. The Charging Pump was used to supply primary system make up water and supply water under high pressure to seal the control rod drive openings in the reactor vessel head Based on CAD drawings, the deck surface area is approximately 15.17 square meters (l 5. l 7m2). The total surface area.including the bulkheads, raised solid deck plating and overhead is 71.14 m2. See Figure 3-2.
3.2.10 STR-109 Auxiliary Access Trunk, C-Deck, Cold Water Chemistry Lab (Port) and Radiation Monitoring Room (Stbd)
STR-109 is an impacted Class I survey unit. On C-Deck, theCold-WaterChemistryLab (Port) and Radiation Monitoring Room (Stbd) were utilized for monitoring water chemistry as well as monitoring air from the RC and CV. At the forward end of the space on centerline, the Auxiliary Access trunk extends upward to A Deck, where it is fitted with a watertight hatch. The survey unit consists of the internal surfaces of the Auxiliary Access Trunk, the Cold-Water Chemistry Lab and the Radiation Monitoring Room. Based on CAD drawings, the deck surf ace area is approximately 30.32 square meters (30.32 m2); the total surface area, including the bulkheads and overhead, is approximately 192.10 m2 ; this includes the surfaces above 2 meters. See Figure 3-2.
Rev. I 27
Final Status Survey Final Report 1 of 5, Revision 1 3.2.11 STR-207 Health Physics Lab STR-207 is an impacted Class 2 survey unit. The HP Laboratory was utilized primarily as a count room and for processing film badges. Records also indicate that decontamination of equipment was occasionally performed in the Health Physics Lab on A Deck. The survey unit consists of base and wall cabinets, countertop and sink. The deck surface area is approximately 13.83 square meters (13.83 m2), based on CAD drawings. The total surface area including the bulkheads and overhead is 73.51 m2. See Figure 3-2.
3.2.12 STR-301 Navigation Bridge Deck - interior surfaces STR-301 is an impacted Class 3 survey unit. The Navigation Bridge Deck contains the pilot house, gyro, radio and chart rooms. It also has fan rooms, battery rooms and an emergency generator room Based on CAD drawings, the deck sutface area is approximately 290.55 square meters (290.55 m2).
Thetotalsurfacearea,includingthebulkheads and overhead, is approximately 1,574.69 m 2. See Figures 3-2 and 3-3.
3.2.13 SYS-112 Primary Pressurizing System (PE) including retained portions of the pressurizer The Primary Pressurizing (PE) System, including the Pressurizer itself, is described in the Updated Final Safety Analysis Report, Rev 13, Reference ( e ). It states that all of the PE system equipment is dismantled and disposed, and that the Pressurizer itself is dismantled and disposed except for its upper shell and support skirt. This survey unit, designated SYS-112, is limited to the retained portions of the Pressurizer shell and skirt. The pressurizer shell has an inside diameter of 4.5 feet and stands approximately 19 feet tall. Dismantlement and decontamination of the pressurizer began in 2022. A large manway was cut into the port side of the shell, through which access was gained to remove all internal components, and cut free the hemispherical bottom portion of the shell containing the heater wells. The interior surface of the shell was decontaminated using a combination of techniques, including laser ablation and mechanical grinding. See Figure 3-4.
3.2.14 SYS-117-01 Neutron Shield Tank wall located in the CV SYS-11 7-0 I is an impacted systems survey unit. The neutron shield tank was a primary shield filled with water that surrounded the reactor vessel. The survey unit consists of the interior and exterior surfaces of the retained outer annular tank wall. The total surface area is approximately 148 square meters (148 m2), based on CAD drawings. See Figure 3-4.
3.2.15 SYS-117-02 Fuel Transfer Tank wall located in the CV SYS-117-02 is an impacted system survey unit. The Fuel TransferTanksurrounds the reactor vessel head above the hot legs. It was used to reduce dose during refueling. The Fuel Transfer Tank is a single tank walJ that is an extension of the Neutron Shield Tank outer annular wall that reaches above the RPV flange. The survey unit consists of the interior and exterior surf aces of that single tank wall. The total surface area is approximately 62 square meters (62 m2), based on CAD drawings. See Figure 3-4.
Remainder of page intentionally blank.
Rev. 1 28
Final Status Survey Final Report 1 of 5, Revision 1 3.3 Locations on General A"angement Drawings Figure 3-2 Inboard Profile r
l Fourteen of fifteen SURRs are located in red rectangle (see SURRs for details):
7
...... -H
........ ~
l r,. -r*
7 The fifteenth STR-301, Navigation Bridge Deck - interior surfaces. See the oval above and in detail below.
Figure3-3 STR-301, Navigation Bridge Deck fj~
.. mi, ~~
See SURR for details.
Rev. I
- r
~ y-.
29
Figure3-S Final Status Survey Final Report 1 of S, Revision 1 STR-105-01, -02 aod-03, Reactor Compartment-Lower Level t4" 01A. AaZIS -
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- * ;........! TT T~--
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f
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~
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Rev. I N. S. "SAVANNAH" ONE POINT VIEW OF SECONDARY SHIELD INTERIOR 5(01)NDMY SHl8.D NOTE :
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UDOERS AHO PUTI"ORMS 31
Final Status Survey Final Report 1 of 5, Revision 1 4
SUMMARY
OF SURVEY RESULTS 4.1 Total Measurements Collected in Each Survey Unit Table 4-1 presents a summary of all the static measurements. The systematic measurements for each Class 1 and 2 survey unit are shown in Column 2. For Class 3 survey units, the measurement locations were selected based on a random grid with a random starting point and shown in Column 3. The judgmental measurements are biased measurements perfonned in locations based on site knowledge and professional judgment to provide added assurance that residual contamination on the ship has been adequately located and characterized. The replicate measurements are QC measurements taken by a different technician or with a different instrument. The last column, # additional measurements, is the total number of measurements made in addition to the number of measurements specified in the FSS package. This number can be from more than one (1) of the following categories; confirmatory measurements taken as a result of exceeding the scanning action level, post remediation measurements, and additional judgmental measurements as directed by the field supervisor.
Remainder of page intentionally blank.
Rev. 1 32
Final Status Survey Final Report 1 of 5, Revision 1 Table 4-1 Summary of Static Measurements Survey Unit
- Systematic
- Random
- Judgmental
- QC Replicate
- Additional Number Measurements Measurements Measurements Measurements Measurements STR-101-01 15 0
2 1
24 STR-101-02 15 0
2 1
6 STR-102 15 0
2 1
4 STR-103 15 0
2 1
0 STR-104 15 0
2 1
0 STR-105-01 15 0
2 1
0 STR-105-02 15 0
2 1
0 STR-105-03 15 0
2 l
0 STR-108 15 0
2 1
0 STR-109 15 0
2 1
0 SYS-112 17 0
2 1
2 SYS-117-01 15 0
2 1
0 SYS-117-02 15 0
2 1
0 STR-207 15 0
2 1
0 STR-301 0
17 2
1 0
Rev. 1 33
Final Status Survey Final Report 1 of 5, Revision 1 4.2 Summary of Survey Results The maximum activity concentration measured of all the final systematic, random and judgmental static measurements was 5.07E+o4 dpm./100 cm2. This value is 72. 7% of the gross activity DCGLw. Table 4-2 presents a summary of all the systematic, random and judgmental static measurement parameters.
Table4-2 Summary of Random and Static Measurement Parameters Parameter STR-101-STR-STR-102 STR-103 STR-104 Units 01 101-02 Minimum Measurement 8.28E+-02 1.08E-+-03 6.03E+-02 5.73E-+-02 l.06E-+-03 dpm/100cm2 Maximum Measurement 3.77E+-04 2.45E-+-04 2.49E+-04 9.20E-+-03 5.07E-+-04 dpm/100cm2 Mean 9.53E+-03 5.18E-+-03 4.07E+-03 l.82E-+-03 4.91E-+-03 dpm/100cm2 Median 3.46E+-03 2.13E-+-03 8.54E+-02 7.74E-+-02 l.35E-+03 dpm/100cm2 Standard Deviation l.06E+o4 5.96E-+-03 6.38E+-03 2.52E-+-03 l.19E-+-04 dpm/100cm2 Coefficient of Variation 1.12E+o0 l.15E-+-OO 1.57E+-O0 l.38E-+-OO 2.42E-+-OO Unitless Parameter STR-105-STR-STR-105-STR-108 STR-109 Units 01 105-02 03 Minimum Measurement 2.87E+-02 3.49E-+-02 5.09E+-02 4.02E-+-02 3.55E-+-02 dpm/100cm2 Maximum Measurement l. l 5E+-03 l.47E-+-03 3.93E+-03 l.22E-+-03 5.61E-+-02 dpm/100cm2 Mean 5.77E+o2 6.3 lE-+-02 1.28E+-03 7.00E-+-02 4.59E-+-02 dpm/100cm2 Median 5.41E+-02 4.97E-+-02 7.80E+-02 6.53E-+-02 4.41E-+-02 dpm/100cm2 Standard Deviation 1.89E+-02 3.00E-+-02 1.08E+-03 1.81E-+-02 6.93E-+-01 dpm/100cm2 Coefficient of Variation 3.30E-01 4.80E-01 8.S0E-01 2.60E-0l l.S0E-01 Unitless Parameter SYS-112 SYS-SYS-STR-207 STR-301 Units 117-01 117-02 Minimum Measurement 2.52E+-02 3.20E-+-02 8.83E+-02 4.95E-+-02 3.72E-+-02 dpm/100 cm2 Maximum Measurement 2.92E+-03 6.26E-+-03 1.85E+-04 8.74E-+-02 1.04E-+-03 dpm/100 cm2 Mean 6.64E+-02 1.72E-+-03 5.43E+-03 6.84E-+-02 5.93E-+-02 dpm/100 cm2 Median 3.66E+-02 1.1 0E-+-03 2.14E+-03 6.59E-+-02 5.72E-+-02 dpm/100 cm2 Standard Deviation 6.91E+-02 2.00E-+-03 5.92E+-03 l.22E-+02 1.28E-+-02 dpm/100cm2 Coefficient of Variation 1.04E+-O0
- 1. l 6E-+-OO 1.09E+-O0 1.80E-0l 2.20E-01 Unitless Rev. 1 34
Final Status Survey Final Report 1 of 5, Revision 1 4.3 Anomalous Data, Elevated Scan Results and Investigations No anomalous data was encountered as the result of the analysis of the collected data.
Fourteen locations triggered scan levels. Five of the fourteen also triggered Investigation Levels equal to the operational DCGLw of 5.23E+o4 dpm/100 cm2. As described in the following sections, all were remediated and resurveyed to confirm they are less than the operational DCGLw.
4.3. I STR-101-01, Containment Vessel (CV) - 1st Level (Tanktop), Starboard Side Eleven locations triggered the scan action level of 2,500 cpm, requiring a confirmatory static measurement at each location. None of these locations were part of the fifteen systematic/unbiased measurement locations.
Three of those eleven static measurements also triggered the Investigation Level equal to the operational DCGLw of 5.23E+o4 dpm/100 cm2. The activity concentration at SA-01-02-J was 9.89E+o4 dpm/100 cm2, SA-01-03-J was 6.76E+o4 dpm/100 cm 2, and SA-01-12-J was 6.00E-+-04 dpm/100 cm 2. The results for the remaining measurements ranged from 1,131.12 dpm/100 cm 2 to 37,716.98 dpm/100 cm2.
Remediation was performed in four 3 ft by 1 ft areas that encompassed nine of the eleven locations exceeding the 2,500 cpm scan action level. Two locations were inaccessible for decon.
The post remediation measurements of the three that triggered the Investigation Level are as follows:
For SA-01-02-J, it is 2.18E+o4 dpm/100 cm2.
For SA-01-03-J, it is 1.49E+o3 dpm/100 cm2.
For SA-01-12-J, it is l.75E+o4 dpm/100 cm2.
All post remediation static measurements are less than the operational DCGLw.
4.3.2 STR-101-02, Containment Vessel (CV)- 1st Level (Tanktop), Port Side One location triggered the scan action level of 2,500 cpm, requiring a confirmatory static measurement. The measurement at SA-01-01 is 50,725 (dpm/100 cm2) which is below the investigation level.
One static measurement location, STR-101-02-16-J triggered the Investigation Level equal to the operational DCGLw of 5.23E+o4 dpm/100 cm2. Its activity concentration was l.49E+o5 dpm/100 cm2* This location was not part of the fifteen systematic/unbiased measurement locations.
Remediation was performed. The post remediation measurement for STR-101-02-16-J is 2.45E+o4 dpm/100 cm2.
The post remediation static measurement is less than the operational DCGLw.
4.3.3 STR-102, Containment Vessel (CV)-2nd Level (Flat)
Two locations triggered the scan action level of 2,500 cpm, requiring a confirmatory static measurement at each location. One of those two static measurements, SA-01-02, also triggered the investigation level equal to the Operational DCGLw of 5.23E+o4 dpm/100 cm2. Its activity concentration was 7.59E+o4 dpm/100 cm2. The other location, SA-01-01, had an activity concentrationof329E-+-04dpm/100 cm2. Neitherofthesetwo (2) locations were part of the fifteen systematic/unbiased measurement locations.
Rev. 1 35
Final Status Survey Final Report 1 of 5, Revision 1 Remediation was performed on both locations. The post remediation measurement for SA-01-1 is l.43E-04 dpm/100 cm2 and SA-01-02 is l.69E+o4 dpm/100 cm2.
All post remediation static measurements are less than the operational DCGLw.
4.3.4 SYS-112, Primary Pressurizing System (PE)
Two locations triggered the scan action level of 1,500 cpm, requiring a confirmatory static measurement at each location. The results of the static measurements are SA-01-01 equal to 12,018.18 dpm/100 cm2, and SA-01-02 equal to 8,527.19 dpm/100 cm2. Neither measurement triggered the Investigation Level of 17,700 dpm/100 cm2.
4.4 Comparison of Findings with the Release Criteria The dose contribution from residual activity in each survey unit was calculated. The mean activity from all the final static measurements was divided by the gross activity DCGL to derive a mean fraction. The mean fraction was then multiplied by 15 mrem/yr to derive the dose attributed to the survey unit.
Table 4-3 presents the results of the calculations.
Table4-3 Residual Activity Dose Rates in Each Survey Unit Survey Unit Dose Rate (mrem/yr)
STR-101-01 2.05 STR-101-02 1.12 STR-102 0.88 STR-103 0.39 STR-104 1.06 STR-105-01 0.12 STR-105-02 0.14 STR-105-03 0.27 STR-108 0.15 STR-109 0.10 SYS-112 0.42 SYS-117-01 0.37 SYS-117-02 1.17 STR-207 0.15 STR-301 0.13 All fifteen survey units meet the release criteria and are acceptable for unrestricted use.
The maximum dose contribution from residual activity in all fifteen survey units is 2.05 mrem/yr TEDE.
This value is from STR-101-01, Containment Vessel (CV) - 1st Level (Tanktop ), Starboard Side.
Rev. 1 36
Final Status Survey Final Report 1 of S, Revision 1 4.5 NRC and Other Independent Verif,cation Findings Inspections are an important element ofNRC's oversight of MARAD. The NRC conducts inspections to ensure that MARAD meets NRC's regulatory requirements. As stated on their website, "When licensees meet these requirements, we know that they are most likely conducting safe operations that protect the public and the environment from any undue nuclear risk."
NRC conducts inspections of licensed nuclear power plants, fuel cycle facilities, and radioactive materials activities and operations. Inspectors follow guidance in the NRC Inspection Manual, which contains objectives and procedures to use for each type of inspection. If an inspection were to show that MARAD is not safely conducting an activity, the NRC would notify MARAD of the concern and monitor its resolution until the issue was resolved the NRC's satisfaction. Follow-on inspections would include monitoring MARAD to ensure the issue did not recur. MARAD has had no inspection violations or findings since February of 2001.
The scope of inspections includes the organizational structure, design, maintenance, and environmental and radiation protection programs to ensure they are adequate and comply with NRC safety requirements.
To support inspection planning of FSS and waste shipment activities, MARAD routinely provided near-term look-ahead schedules of these activities to the NRC inspector. As documented in NRC Inspection Report No. 05000238/2024001 (Reference u), the NRC inspected RASS and FSS activities in 2024.
Significant NSS FSS inspections by NRC include the following:
On March 7, 2024, NRC observed RASSs.
On June 5, 2024, NRC observed FSS activities in the CV.
On November 7, 2024, NRC observed an FSS walkdown of the FWST the daily 1400 FSS team meeting.
The most significant inspection of MARAD FSS activities was performed July 29 - August 1, 2024.
During this period, the Oak Ridge Institute for Science and Education (ORI SE) conducted confirmatory surveys as an independent entity from NRC and MARAD to assess the adequacy of the licensee's final status survey results. NRC inspectors observed these surveys. The NRC relies, in part, on the confirmatory surveys to assess the adequacy of licensee surveys to demonstrate residual radioactivity levels meet the release criteria. The following observations, reviews, walkdowns, interviews and determinations were documented in the inspection report:
Inspectors identified areas to be surveyed and sampled, worked with ORISE to determine what type of survey would be conducted, and developed a plan to perform measurements and sample collection.
NRC inspectors identified locations where NSS technicians collected samples for the confirmatory survey.
Inspectors observed site personnel conducting FSSs.
Inspectors observed gamma scanning measurements and the collection of metallurgical samples at various locations within the CV.
Inspectors conducted interviews with cognizant site personnel to determine if FSS activities were conducted in accordance with the site requirements.
Inspectors observed performance of RASSs and FSSs.
Inspectors reviewed instruments used in support of RASS and FSS activities and observed instrument source checking.
Rev. 1 37
Final Status Survey Final Report 1 of 5, Revision 1 Inspectors reviewed documentation that included procedures, survey records, and the license termination plan.
Inspectors reviewed dosimetry including environmental.
Inspectors walked down areas of the ship after FSS was completed to observe isolation controls to prevent unwanted access.
Inspectors reviewed training and qualifications of personnel performing RASS and FSS.
Inspectors reviewed documentation that included FSSs, overview of MARRSIM and license termination process.
Inspectors determined that the procedures that govern RASS and FSS were appropriate and implemented in a manner that provides reasonable assurance the site does not pose undue risk to public health and safety. This included chain of custody, access control to FSS areas, survey data collection and data management, survey quality assurance requirements, and records retention requirements.
Inspectors determined radiological instruments used in support of RASS and FSS were appropriate.
Their selection and calibration were appropriate for the ROC and the DCGLs.
Inspectors determined the NSS did not have any environmental releases. However, environmental results were reviewed for both water and sediment Inspectors reviewed area occupational radiological exposure results. The review of these results were below regulatoiy limits, and vendors used to analyze samples and dosimetry were determined to be accredited.
Inspectors determined that the licensee developed survey and sampling plans that were appropriate to the area remediated consistent with the site's classification of the survey unit. Additionally, the required percentage of the area to be surveyed was adequate, and the method to determine background radioactivity was appropriate.
Inspectors determined the quality controls of assessing and reporting of survey results were appropriate.
Inspectors determined individuals tasked with performing surveys were qualified to conduct surveys for RASS and FSS.
Inspectors determined these surveys were performed in accordance with applicable procedures and plans.
All inspector questions were answered on the spot or prior to the Exit Meeting. No follow up action items were identified.
In July 2025, ORISE documented their activities in Reference (v).
MARAD maintains two contracts that provide independent review and oversight of its licensed activities, and field inspection and regulatory compliance services. These contracts are held by B2Z Engineering3 and Sustainable Design Consortium respectively and are described in STS-222, Decommissioning Funds Status Report for CY2022, Reference (w). Both companies conducted periodic independent assessments of decommissioning activities, including the planning and execution of the Final Status Survey program.
3 The B22 contract expired on September 30, 2025. The SDC contract is expected to remain in place through license termination.
Rev. 1 38
Final Status Survey Final Report 1 of 5, Revision 1 The assessments included observing surveys in progress, interviewing site personnel and review of FSS packages and SURRs. Site visits were performed in July of 2023 by B2Z Engineering and in August of 2024 by Sustainable Design Consortium via Tidewater, Inc.
In July 2024, MARAD approved CR-174, Metallurgical Sampling FSS Program, (Reference x) for collecting samples to validate the radionuclide mix proposed for the DCGLs and evaluate neutron activation of systems and components that would remain in place after termination. Report CR-176, Metallurgical Samples Evaluation, (Reference y) documents the evaluation of those samples. It concluded that the DCGLs and RO Cs listed in the L TP are valid, It also concluded that the remaining systems do not indicate neutron activation.
5 FSS FINAL REPORT
SUMMARY
All fifteen (15) survey units attached to this report are acceptable for unrestricted use, based on the following findings:
All fifteen survey units met the DQOs of the survey plans. This demonstrates that the survey units have been properly classified.
The maximum activity concentration observed in all final static measurements was 5.07E+o4 dpm/100 cm2. This value is 72. 7% of the gross activity DCGLw.
The sample data from all fifteen survey units passed the Sign Test. The null hypothesis was rejected. Therefore, all fifteen survey units meet the release criteria.
The maximum dose contribution from residual activity in all fifteen survey units is 2. 0 5 mrem/yr TEDE. This value is based on the mean activity of all final static measurements No anomalous data was encountered as the result of the analysis of the collected data. Fourteen locations triggered scan levels. Five of the fourteen also triggered Investigation Levels equal to the operational DCGLw. All were remediated and resurveyed to confirm they are less than the operational DCGLw.
Isolation and Control procedures continue in effect, and to date, no new residual radioactivity has been discovered in any unit.
6 References
- a.
STS-004-003, NS. SAVANNAH License Termination Plan, Revision 1
- b.
NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM),
Revision 1, August 2000 c
NUREG-17 57, Vol 2, Revision 1, Characterization, Survey, and Determination of Radiological Criteria
- d.
STS-005-035, Preparation of Survey Unit Release Records and FSS Final Reports
- e.
STS-003-001, Decommissioning Quality Assurance Plan
- f.
STS-005-029, Final Status Survey Program
- g.
STS-005-030, Preparation of Final Status Survey Packages
- h.
STS-005-031, Calculation of The Number of Measurements For Final Status Surveys
- i.
STS-005-032, Survey Unit Turnover and Control Rev. 1 39
Final Status Survey Final Report 1 of S, Revision 1
- j.
STS-005-033, Final Status Survey Data Assessment and Investigation
- k.
STS-005-034, Survey Unit Classification I.
SIC-MA-Q-25, FSS Survey Unit Inspections After Surveys are Completed
- m.
Visual Sample Plan, Pacific Northwest National Lab, USDOE
- n.
CR-109, RSCS TSD 19-031, Radiological Characterization - Reactor Compartment and Containment Vessel, Revision 1, February 2020
- o.
CR-139, TSD 21-089, Calculations to Support NSS Surface Contamination DCGLs, Revision O 1, September 2022
- p.
NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey for Instruments for Various Contaminants and Field Conditions, Revision 1, 2020
- q.
CR-164, TSD 23-126, Calculation of Weighted Efficiencies and MDCs of Ludlum Detectors for Final Status Surveys on the NS SAVANNAH, Revision 01
- r.
CR-003, STS-109, Historical Site Assessment, 2023.
- s.
NUREG-1576,Multi-Agency Radiological Laboratory Analytical Protocols Manual, MARLAP, July 2004.
- t.
STS-002-001, STS Training Program
- u.
Letter from Elise Eve (NRC) to Mr. Erhard W. Koehler (MA.RAD), dated March 20, 2025, US.
Department of Transportation, N.S. SAVANNAH -NRC Inspection Report No. 05000238/2024001
- v.
Letter from Erika N. Bailey (ORISE) to Tanya Hood (NRC), dated July 7, 2025, CONFIRMATORY SURVEY OF THE NUCLEAR SHIP SAVANNAH LOCATED IN BALTIMORE, MARYLAND DOCKET NUMBER 05000238; RFTA 24-003; DCN 5381-SR-01-0
- w.
STS-222, Decommissioning Funds Status Report/or CY2022, March 2023
- x.
CR-174, Metallurgical Sampling FSS Program, July 2024
- y.
CR-176, Metallurgical Samples Evaluation, December 2024 Rev. 1 40
Final Status Survey Final Report 1 of S, Revision 1 7 ATTACHMENTS 7.1 STR-101-01 Containment Vessel (CV)- JS' Level (Tanktop), Starboard Side 7.2 STR-101-02 Containment Vessel (CV)- 1st Level (Tanktop), Port Side 7.3 STR-102 Containment Vessel (CV)-2nd Level (Flat) 7.4 STR-103 Containment Vessel (CV)-3rd Level (D Deck) 7.5 STR-104 Containment Vessel (CV)-4th Level (C Deck) 7.6 STR-105-01 Reactor Compartment-Lower Level (5'-23') Starboard Half
- 7. 7 STR-105-02 Reactor Compartment-Lower Level (5 '-23 ') Port-side Half 7.8 STR-105-03 Reactor Compartment-Lower Level (5'-23') Drain Wells 7.9 STR-108 Starboard Charging Pump Room 7.10 STR-109 Auxiliary Access Trunk, C-Deck, The Cold Water Chemistry Lab (Port) and Radiation Monitoring Room (Stbd) 7.11 STR-207 Health Physics Lab 7.12 STR-301 Navigation Bridge Deck - interior surfaces 7.13 SYS-112 Primary Pressurizing System (PE) including retained portions of the Pressurizer 7.14 SYS-117-01 Neutron Shield Tank wall located in the CV 7.15 SYS-117-02 Fuel Transfer Tank wall located in the CV Rev. 1 41