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U. S. Dept. of Transportation - Submittal of Final Safety Analysis Report, Revision V
	ML091280436
Person / Time
Site:	NS Savannah
Issue date:	05/04/2009
From:	Koehler E; US Dept of Transportation (DOT)
To:	; Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
	Download: ML091280436 (166)
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0 U.S. Department of Savannah Technical Staff 1200 New Jersey Ave., SE Transportation Office of Ship Disposal Washington, DC 20590 Maritime Administration Ref: 10 CFR 50.71(e)

May 4, 2009 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

Docket No. 50-238; License No. NS-1; N.S. Savannah Submittal of Final Safety Analysis Report, Revision V Pursuant to 10 CFR 50.71 (e), the Maritime Administration (MARAD) hereby submits Revision V of the Final Safety Analysis Report as Enclosure (1). MARAD certifies that the updated information accurately presents changes made since the previous revision.

Revision V is resubmitted in its entirety and not on a replacement-page basis. The updated FSAR includes the effects of the following:

all changes made in the facility or procedures as described in the FSAR;

" all safety evaluations performed to support a license amendments; and,

" all analyses of new safety issues performed by or on behalf of the licensee at Commission request.

This submittal contains no new Regulatory Commitments.

If there are any questions or concerns with any issue discussed in this request, please contact me at (202) 366-2631, and/or e-mail me at erhard.koehler~dot.gov.

Respect Erhard W. Koehler Senior Technical Advisor, N.S. Savannah Office of Ship Disposal Enclosure

Docket No. 50-238; License NS-1; N.S. Savannah Submittal of Final Safety Analysis Report, Revision V May 4,2009

Enclosure:

1. Final Safety Analysis Report, Revision V 2

Docket No. 50-238; License NS-1; N.S. Savannah Submittal of Final Safety Analysis Report, Revision V May 4,2009 cc:

Electronic copy NSS ESC NSS SRC Division of Atlantic Operations MAR 615 Hardcopy, cover letter only MAR-600, 640, 640.2 Hardcopy w/ all enclosures and attachments MAR-100, 640.2 (rf)

USNRC (John T. Buckley, Mark C. Roberts)

USNRC Regional Administrator - NRC Region I MD Department of the Environment (Roland G. Fletcher; George S. Abum, Jr.)

NC Department of Environment & Natural Resources (Beverly 0. Hall)

SC Department of Health & Environmental Control (Susan E. Jenkins; Aaron Gantt)

VA Department of Emergency Management (Michael M. Cline)

VA Department of Health (Leslie P. Foldesi)

EK/jmo 3

0 U.S. Department of Savannah Technical Staff Office of Ship Disposal 1200 New Jersey Ave., SE Washington, DC 20590 Transportation Maritime Administration Docket No. 50-238; License No. NS-1; N.S. Savannah to Submittal of Final Safety Analysis Report, Revision V I

U.S. Department of Transportation Maritime Administration Office of Ship Disposal N.S. Savannah UPDATED FINAL SAFETY ANALYSIS REPORT

-STS-004-002 Revision V Approved: Ma/6j'~yi 2009 Senior Technical Advisor Date Prepared by:

Sayres and Associates Corporation

Updated Final Safety Analysis Report - (STS-004-002)

RECORD OF REVISIONS Summary of Revisions Revision Following Final Shutdown and Permanent Defueling STS-004-002, A complete revision to the FSAR that incorporates numerous changes that Revsi Ioccurred from 1968 through March 2007 Revision IV STS-004-002, These revision numbers to the STS were not used. By not using these numbers, the revision number for the FSAR and the revision number to STS-004-002 will be Revisions 1-4 the same revision number.

STS-004-002, This revision incorporates changes that occurred after March 2007. The most Rev. 5; significant changes are in Chapter 2 (Fire Detection and Alarm Systems, Security FSAR Revision Systems and General Alarm) and Chapter 3 (Ship Movement and Port Operating V Criteria)

PREVIOUS ISSUES N.S. Savannah SAFETY ASSESSMENT, Volumes I-V, 1960-1962 N.S. Savannah HAZARDS

SUMMARY

REPORT FOR DOCKSIDE OPERATION, September 1963 N.S. Savannah HAZARDS

SUMMARY

REPORT FOR 1964 SEA TRIALS, December 1963 N.S. Savannah SAFETY ASSESSMENT, Revision I, July 1965 N.S. Savannah SAFETY ASSESSMENT, Revision II, March 1967 SAFETY ANALYSIS REPORT N.S. Savannah FUEL SHUFFLE, December 1, 1967 N.S. Savannah SAFETY ASSESSMENT, Revision 111, October 1968 SAFETY ANALYSIS

SUMMARY

- N.S. Savannah DEFUELING, February 1971 SUPPLEMENT NO. I TO SAFETY ANALYSIS REPORT N.S. Savannah FUEL SHUFFLE -

FUELING WITH CORE II, October 1971 (Note that Core II was never installed)

UPDATED FINAL SAFETY ANALYSIS REPORT, Revision IV, April 2007 Rev. V

ii

Updated Final Safety Analysis Report - (STS-004-002)

LIST OF EFFECTIVE PAGES Page Rev. Page Rev. Page Rev. Page, Rev. Page Rev.

No. No. "No. No. No. No. No >.N. No. No.

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1122 125 5 Rev. V ,°,

Updated Final Safety Analysis Report - (STS-004-002)

Page Rev. Page Rev. Page Rev. Page. "Rev. Page Rev.

No. No. -.No.. No. No. No. No. No. No. No.

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- )' : *: : :' : :' i ii Rev. V iv

Updated Final Safety Analysis Report - (STS-004-002)

Table of Contents Record of Revisions ...................................................................................................................................... ii Previous Issues .............................................................................................................................................. ii List of Effective Pages ................................................................................................................................. iii Foreword to Revision V (2009) ................................................................................................................. xiii Foreword to Revision IV ........................................................................................................................... xiii I INTRODUCTION AND SUM M ARY ............................................................................................. 1 1.1 Sum m ary of Significant Events .............................................................................................. I 1.2 Principal Safety Features of the Ship ....................................................................................... 8 1.2.1 The Hull ..................................................................................................................................... 8 1.2.2 The Power Plant ............................................................................................................... 9 1.3 Ship and System Status - 1976 to Present .............................................................................. 10 1.3.1 Power Plant Status as reported in April 1976 .......................................... ................................ 10 1.3.2 Status of N .S. Savannah from April 1976 - April 2007 ...................................................... 12 1.3.3 Radiological and Environmental Characterization Scoping Survey .................................. 14 1.3.4 Current Status of N .S. Savannah April 2007 to present ...................................................... 14 1.4 Symbols for the Reactor Plant A uxiliary System s ................................................................ 17 2 SHIP DESCRIPTION ..................................................................................................................... 18 2.1 Features and Characteristics .................................................................................................. 18 2.2 Com partm entation, Decks and Space Arrangem ent ............................................................... 18 2.2.1 Com partm entation .................................................................................................................. 18 2.2.2 Deck Arrangem ents ........................................................................................................... 19 2.2.3 Space Arrangem ent ............................................................................................................ 20 2.3 N avigation and Steering Equipm ent ......................... ............................................................ 20 2.3.1 N avigation Equipm ent ........................................................................................................ 20 2.3.2 Steering Gear ............................................................................... I........................................... 20 2.4 Fire Detection and Control System .............................................................................................. 20 2.4.1 Fire Detecting System ................................. ...... 20 2.4.2 Fire Detection and A larm System (FA) .............................................................................. 21 2.5 Lifesaving Equipm ent ............................................................................................................. 21 2.6 Radio Equipm ent ....................................................................................................................... 21 2.7 Deck M achinery ......................................................................................................................... 21 2.8 Flooding A larm s ......................................................................................................................... 21 2.9 Dehum idification System ..................................................................................................... 22 2.10 Hull Cathodic Protection System ......................................................................................... 22 Rev. V v

Updated Final Safety Analysis Report - (STS-004-002) 2.11 Security System s ......................................................................................................................... 22 2.11.1 Camera Surveillance System ............................................................................................. 22 2.11.2 Security Entrance Door ..................................................................................................... 22 2.12 General Alarm ............................................................................................................................ 22 3 SITE AN D SITE CRITERIA ..................................................................................................... 26 3.1 Site Description ......................................................................................................................... 26 3.2 Ship M ovem ent and Port Operating Criteria ......................................................................... 27 3.2.1 Purpose .................................................................................................................................... 27 3.2.2 Ship M ovem ent Criteria ..................................................................................................... 27 3.2.3 Port Operating Criteria ........................................................................................................ 28 3.3 Control Zones ............................................................................................................................. 28 3.4 Comparison with Central Station Site criteria in the USA ........................ 28 4 CON TA INM EN T ........................................................................................................................... 29 4.1 Introduction ................................................................................................................................ 29 4.2 Prim ary Containm ent (Containm ent Vessel) ......................................................................... 29 4.2.1 Function and General Design Criteria .......................... I..................................................... 29 4.2.2 Containm ent Vessel ............................................................................................................ 30 4.2.3 Containm ent Vessel Penetrations ............................................................................. 32 4.2.4 Containm ent A uxiliary Features ......................................................................................... 33 4.2.5 Containm ent Perform ance Requirem ents ........................................................................... 35 4.3 Reactor Com partment (Secondary Containm ent) .................................................................. 35 4.3.1 General Description ............................................................................................................. 35 4.3.2 Cooling Section ........................................................................................................................ 36 4.3.3 Exhaust Section ....................................................................................................................... 36 5 RADIATION SH IELD IN G ....................................................................................................... 52 5.1 Design Basis .............................................................................................................................. 52 5.2 Radiation Sources ....................................................................................................................... 52 5.3 Radiation Shields ........................................................................................................................ 52 5.3.1 Prim ary Shield ........................................................ ................................................................. 52 5.3.2 Secondary Shield .................................................................................................................... 52 5.3.3 Local Shielding........................................................................................................................ 54 5.4 Radiation Sources ....................................................................................................................... 54 5.5 Shield Surveys ............................................................................................................................ 54 5.6 Calculated Doses from Failed Fuel Rods ............................................................................. 54 6 REACTOR CORE .......................................................................................................................... 55 6.1 Reactor Core General Description .......................................................................................... 55 6.1.1 Fuel Elem ents .......................................................................................................................... 55 6.1.2 Control Rods ............................................................................................................................ 55 6.1.3 Internals ................................................................................................................................... 55 Rev. V vi

Updated Final Safety Analysis Report - (STS-004-002) 6.2 Therm al and Hydraulic Characteristics .................................................................................. 57 6.3 N uclear Characteristics ........................................................................................................ 57 6.4 Shipboard Physics Test Program ................................. :.......................................................... 57 7 PRIM ARY COOLAN T SY STEM ............................................................................................ 59 7.1 General D escription .................................................................................................................... 59 7.1.1 General Design Data .......................................................................................................... 59 7.1.2 General Perform ance Data ................................................................................................ 60 7.2 Reactor V essel ............................................................................................................................ 63 7.3 Steam Generators ........................................................................................................................ 65 7.4 Pressurizer .................................................................................................................................. 66 7.5 Prim ary Pumps ........................................................................................................................... 68 7.6 Valves ......................................................................................................................................... 68 7.7 Prim ary (N eutron) Shield Tank ............................................................................................ 68 7.8 Prim ary Piping ............................................................................................................................ 68 8 POW ER CON VERSION SY STEM S ................................. ..................................................... 70 8.1 Steam Generators ........................................................................................................................ 70 8.1.1 General Description ............................................................................................................ 70 8.1.2 Design Conditions ................................................................................................................... 71 8.1.3 Construction Features ........................................................................................................ 72 8.2 M ain and A uxiliary Steam System s ...................................................................................... 72 8.2.1 M ain Steam System ............................................................................................................ 72 8.2.2 Auxiliary Steam System ..................................................................................................... 73 8.2.3 Interm ediate Pressure Steam System ................................................................................... 73 8.3 M ain Propulsion System ....................................................................................................... . 73 8.3.1 High-Pressure Turbine ....................................................................................................... 73 8.3.2 Low -Pressure and Astern Turbine ..................................................................................... 74 8.3.3 Reduction Gears ...................................................................................................................... 74 8.3.4 Propulsion Shafting and Propeller .................................................................................... 74 8.3.5 M ain Condenser ...................................................................................................................... 74 8.3.6 M ain Condensate and Feedw ater System ........................................................................... 75 8.4 Em ergency Propulsion System .............................................................................................. 76 8.5 Electrical Pow er System s .................................................................................................... .76 8.5.1 Shore Pow er Connection ................................................................................................... 77 8.5.2 Pow er D istribution ............................................................................................................ 77 8.6 A uxiliary Engine Room System s .......................................................................................... 83 8.6.1 Instrum ent A ir System ........................................................................................................ 83 8.6.2 W aste D ilution and Disposal (W D ) System ...................................................................... 84 8.6.3 Engine Room Ventilation System ....................................................................................... 84 8.6.4 Blow dow n Evaporator System ............................................................................................ 85 8.6.5 Fresh W ater System ............................................................................................................. 85 8.7 W ater Chem istry and Control ................................................................................................. 86 Rev. V vii

Updated Final Safety Analysis Report - (STS-004-002) 8.7.1 Prim ary System ....................................................................................................................... 86 8.7.2 Secondary System .............................................................................. ..................................... 87 8.7.3 Laboratory A rrangem ents ................................................................................................... 89 8.7.4 Instrum entation ........................................................................................................................ 89 8.7.5 Operating Experience ....................................................................................................... 90 8.8 Pow er Plant Perform ance .................................................................................................... 90 9 REA CTOR PLA N T A UX ILIA RY SY STEM S ......................................................... : ................... 91 9.1 Reactor Emergency System s ................................................................................................ 91 9.1.1 Emergency Cooling (D K ) System ............................................................................................ 91 9.1.2 Soluble Poison (SP) System .............................................................................................. 92 9.2 Reactor A uxiliary System s ................................................................................................. 92 9.2.1 Prim ary Loop Purification (PP) System ........................................................................... .92 9.2.2 Buffer Seal (SL) System ..................................................................................................... 95 9.2.3 Hydrogen A ddition (H A ) System ............................................................................................ 98 9.2.4 Prim ary Relief (PR) System .............................................................................................. 99 9.2.5 Prim ary Sam pling (SA ) System ............................................................................................ 100 9.2.6 Interm ediate Cooling W ater (CW ) System ........................................................................... 102 9.2.7 Containm ent Cooling (CC) System ....................................................................................... 106 9.2.8 Shutdow n Circulation (SC) System ....................................................................................... 109 9.2.9 Prim ary Pressurizing (PE) System ........................................................................................ 109 9.3 Waste Management Systems ........................................... 111 9.3.1 Equipm ent D rain and W aste Collection (PD) System ...................................................... I.I 9.3.2 G aseous Waste Collection and D isposal (W L) System ......................................................... 114 10 IN STRU MEN TA TION A N D CON TRO L ................................................................................... 117 10.1 G eneral D escription .................................................................................................................. 117 10.2 Plant Control A rrangem ent ....................................................................................................... 117 10.2.1 Introduction ........................................................................................................................... 117 10.2.2 M ain Control Room ............................................................................................................... 117 10.2.3 Hydraulic Pow er Supply Room ............................................................................................. 1 18 10.2.4 G E Cubicle ............................................................................................................................ 118 10.2.5 Em ergency G enerator Room ................................................................................................. 118 10.2.6 Forw ard Control A rea .......................................................................................................... 118 10.2.7 Data Acquisition System ............................................ 118 10.3 Instrum entation ......................................................................................................................... 118 10.3.1 Nuclear Instrumentation System ....................................... 118 10.3.2 Safety System .................................................................................................... 118 10.3.3 N on-nuclear Instrumentation ................................................................................................. 119 10.4 Reactor Control ......................................................................................................................... 119 10.5 Control Rod Drive System ........................................................................................................ 119 10.5.1 System D escription ................................................................................................................. 119 10.5.2 Control Rod D rive M echanism s ............................................................................................ 119 10.5.3 System Status as of M arch 1976 ........................................................................................... 119 10.6 Radiation M onitoring System ................................................................................................... 120 Rev. V viii

Updated Final Safety Analysis Report - (STS-004-002) 11 CONDUCT OF LICEN SED ACTIV ITIES .................................................................................. 121 11.1 Organization .................................................................. :.......................................................... 121 11.2 Training .................................................................................................................................... 122 11.3 Procedures ................................................................................................................................ 122 11.4 Records .................................................................................................................................... 122 11.5 Quality A ssurance Plan ............................................................................................................ 123 11.6 Port Operating Plans ................................................................................................................. 123 11.7 Radiation Protection Plan ......................................................................................................... 123 11.8 Industrial Safety ........................................................................................................................ 124 11.8.1 Occupational Health and Environm ental Control ................................................................. 124 11.8.2 Personal Protection ................................................................................................................ 124 11.8.3 Lifting and Handling Equipm ent ........................................................................................... 124 12 SHIP ACCIDEN TS ...................................................................................................................... 125 12.1 Introduction .............................................................................................................................. 125 12.2 Conventional Ship Safety ......................................................................................................... 125 12.3 Collisions and A llisions ............................................................................................................ 125 12.4 Grounding ................................................................................................................................. 126 12.5 Heavy W eather ......................................................................................................................... 127 12.6 Fire and Explosion .................................................................................................................... 128 12.7 Flooding and Sinking ................................................................................................................ 129 12.8 Salvage ...................................................................................................................................... 129 13 RADIOLOGICAL AND ENVIRONMENTAL CHARACTERIZATION .................................. 130 13.1 Radiological Characterization Scoping Survey ........................................................................ 130 13.1.1 Dose Rates ............................................................................................................................. 131 13.1.2 Radiological Contam ination .................................................................................................. 131 13.1.3 N on-Radiological Areas ........................................................................................................ 131 13.1.4 Radiological Areas .................................................................................. ........ 132 13.1.5 Reactor Com partm ent (Secondary Containm ent) ................................................................. 133 13.1.6 Containm ent Vessel ............................................................................................................... 134 13.1.7 Estim ate of the Quantity and Distribution of Radionuclides ................................................. 134 13.1.8 Calculation of Total Contam ination Inventory ...................................................................... 138 13.1.9 Reactor Pressure Vessel (RPV) ............................................................................................. 140 13.2 Environm ental Characterization (HAZM AT) Scoping Survey ................................................ 141 13.2.1 Asbestos-Containing M aterial ............................................................................................... 142 13.2.2 Lead-Based Paint ................................................................................................................... 142 13.2.3 Liquids ................................................................................................................................... 142 13.2.4 Polychlorinated Biphenyls (PCBs) ........................................................................................ 143 13.2.5 M ercury ................................................................................................................................. 144 13.3 Updates to Know n Radiological Conditions ............................................................................ 144 13.4 References ................................................................................................................................ 144 Rev. V ix

Updated Final Safety Analysis Report - (STS-004-002) 14 ACCIDEN T AN A LYSIS .............................................................................................................. 145 14.1 Overview .................................................................................................................................. 145 14.2 Radiation Exposures during Licensed Activities ...................................................................... 146 14.2.1 Offsite and Unrestricted Area Exposure ................................................................................ 146 14.2.2 Occupational Radiation Exposure ......................................................................................... 146 14.2.3 Radiation Exposure from Radwaste Transportation .............................................................. 146 14.3 Protective Action Guidelines (PAGs) ....................................................................................... 146 14.4 M aterials Handling Accident .................................................................................................... 147 14.5 Sum m ary ................................................................................................................................... 148 Rev. V X

Updated Final Safety Analysis Report - (STS-004-002)

List of Tables Table 1-1 Chronology of Significant Events 2 Table 1-2 Tank Designation 1I Table 2-1 Principal Characteristics of the N.S. Savannah 18 Table 2-2 Compartment Utilization 19 Table 4-1 List of Containment Penetrations 40 Table 6-1 Reactor Internals 57 Table 7-1 Primary System Water Quality 62 Table 7-2 Primary System Radiochemistry 62 Table 8-1 Steam Generator Data 71 Table 8-2 Group Control Centers 79 Table 8-3 450-Volt Power Panels 80 Table 8-4 120-Volt Power Panels 82 Table 8-5 Primary System Water Quality 86 Table 8-6 Steam Generator Water Quality 87 Table 9-1 Liquid Drainage and Storage 111 Table 9-2 Double Bottom Tanks below RC 113 Table 13-1 Non-Radiological Area Summary 132 Table 13-2 Radiological Area Summary 132 Table 13-3 Reactor Compartment Radiological Summary 133 Table 13-4 Containment Vessel Radiological Summary 134 Table 13-5 Area Monitoring Results 136 Table 13-6 Sample Analysis Results 137 Table 13-7 Total Surface Contamination Inventory 138 Table 13-8 Steam Generator Dose Rates (primary system side) 139 Table 13-9 Steam Generator Contamination Levels (primary system side) 139 Table 13-10 Volume of water remaining in the Primary Coolant system 140 Table 13-11 Summary of Radioactivity Present 141 Table 13-12 Summary of Radioactivity Present 141 Table 13-13 Radiological Spaces Liquid Sample Summary 143 Rev. V xi

Updated Final Safety Analysis Report - (STS-004-002)

List of Figures Figure 2-1 Containment, Reactor Compartment and Engine Room (Frames99-148) 23 Figure 2-2 General Arrangement - Inboard Profile and Decks - Promenade, A and B 24 Figure 2-3 General Arrangement - Decks - C, D, Flats, Tank Tops and Containment Vessel 25 Figure 4-1 Arrangement of Principal Components in Containment 30 Figure 4-2 Containment Inerting Schematic 35 Figure 4-3 Reactor Compartment 36 Figure 4-4 Containment Vessel (showing the secondary shield) 39 Figure 5-1 Primary Shield Tank 53 Figure 6-1 Reactor Internals 58 Figure 7-1 Plant Arrangement in the Containment Vessel 61 Figure 7-2 Reactor Vessel during Fabrication 63 Figure 7-3 Reactor Vessel 64 Figure 7-4 Steam Generator 65 Figure 7-5 Pressurizer 67 Figure 9-1 Double Bottom Tanks below RC 114 Figure 9-2 Double Bottom Tanks 116 Rev. V xii

Updated Final Safety Analysis Report - (STS-004-002)

FOREWORD TO REVISION V (2009)

Revision V to the UFSAR is the 2009 biennial update. This revision incorporates changes based on improved staff knowledge and understanding of the condition of the ship and its infrastructure based on continuous crewing since 2007. Revision V also incorporates materials and information contained in license submittals over the past two years, including the most recent (April 2008) license amendment (14), the May 2008 Environmental Assessment (EA), and the December 2008 Post Shutdown Decommissioning Activities Report (PSDAR). Minor formatting changes are included to clarify the intent and structure of the document.

The ship and its nuclear facilities remain in protective storage. The ship is in an active layberthing condition, with a retention crew of four persons and regular attendance by the Maritime Administration's nuclear operations staff. This condition is projected to continue throughout the next two-year biennial UFSAR update period. The updated safety analyses found in this revision apply to all activities within the site boundary (i.e.; all shipboard activities), regardless of how they may be characterized.

The near-term focus of decommissioning planning, as described in the EA and PSDAR, is to bring the ship and its nuclear facilities into conformance with contemporary NRC SAFSTOR criteria, and to place the ship into active retention for a period of as much as 15 years before starting DECON and License Termination. This revision provides appropriate safety analyses and bounding conditions for these activities..

FOREWORD TO REVISION IV Revision IV to the FSAR is intended to provide a 2007 status of the N.S. Savannah. It is a complete rewrite of Revision III and includes a summary of numerous documents submitted to the NRC since the last update in 1968, including the most recent (January 2007) license amendment (13).

At the end of 1972, after the ship had been defueled and placed in wet layup status, the Maritime Administration decided not to refuel the ship and return it to service. In effect, this decision made the 1970 and 1971 actions permanent. The final shutdown was in November 1970 and the permanent defueling was completed in fall 1971, these actions are retroactively considered to be permanent cessation of operations, Therefore, the fall 1971 defueling became the de facto permanent defueling and cessation of operations with the issuance of Technical Specification Change 13.

The ship is in a state of protective storage. Most weather openings are sealed. Almost no mechanical system or equipment is functional excepted as noted in the following text. The functional systems include a significant portion of the electrical distribution system, the hull cathodic protection system, the dehumidification (DH) system, the intrusion, fire and flooding alarm system and the anchor windlass.

The DH system was newly installed in 1994, and is mechanically independent of any other ship's system.

Likewise, the alarm system was substantially renewed in 1994 and is isolated from any other ship's system except electrical distribution. Note that modifications associated with the DH and alarm systems (i.e., vent ducting and cabling) generally obstructs the doorways, passageways and ladders throughout ship as is typical in the marine industry for deactivated ships.

The electrical distribution system, hull cathodic protection system and the anchor windlass are the only currently functional portions of the ship's original outfit. Since 1994, there has been minimal maintenance and testing to the ship's electrical system. Any future use of the electrical system would require surveying prior to energizing any of its currently deenergized portions.

Rev. V Xlll

Updated Final Safety Analysis Report - (STS-004-002) 1 INTRODUCTION AND

SUMMARY

1.1 Summary of SignificantEvents A nuclear merchant ship was first proposed by President Eisenhower in 1955 as evidence of this country's interest in the peaceful uses of atomic energy. In 1956, Congress authorized the program in Public Law 848, 84th Congress, which provided funds for the design and construction of a vessel, subsequently named the N.S. Savannah (NSS).

On October 15, 1956, the President directed the Atomic Energy Commission (AEC) and the Maritime Administration (MARAD) to proceed with design and construction. To complement this effort, the entire project was placed under the direction of a dual status staff chosen jointly by MARAD and the AEC. The vessel was designed by George G. Sharp, Inc. Naval Architects of New York. The Babcock and Wilcox Company (B&W) was awarded a fixed price contract for the design and construction of the nuclear power plant auxiliaries. The New York Shipbuilding Corporation (NYS) was selected to construct and test the ship at their Camden, New Jersey shipyard. In 1958, the States Marine Lines, Inc. was engaged as the operating agent for the ship. Todd Shipyard Corporation was chosen to provide the maintenance facility in May 1960. In 1963, American Export lsbrandtsen Lines Inc. (AEIL) became the operating agent. The initial operating License was issued in August 1965. First Atomic Ship Transport Inc., (FAST) a subsidiary of AEIL, placed the NSS in routine cargo service in September 1965.

The NSS, a vessel of the P-2 class (specifically P2-NI-MA40a) was powered by an 80 MWth pressurized-water nuclear reactor and had a design speed of 22 knots at 22,000 shaft horsepower. The normal cruising speed was 20.25 knots at 20,000 shaft horsepower. The ship was intended to promote international goodwill by being displayed in the seaports of the world. Consequently, it was made as attractive and modern as any vessel of that era and its nuclear power plant was designed with an exceptional degree of reliability and safety.

The prime objectives of the creators of the NSS were to demonstrate the advanced technology and peaceful use of nuclear energy. The operation of a working test-bed laboratory ship provided a tool to be used in the development of economical ships and marine reactor operation. It also established acceptable standards for the design of ship and reactor, operating practices, manning, port entry and operation and safety of crew and the general public. It was necessary therefore, that entry into the domestic and foreign ports was in accordance with mutually acceptable regulations and operational procedures. The successful culmination of these objectives led to the general acceptance of the NSS's safety and reliability by governmental authorities, cargo shippers and the general public.

During its operating phase, regulatory procedures, inspection standards and training requirements were improved and implemented. The relation of manning to the ship's layout, design and service was studied.

The high degree of safety, reliability and performance of the NSS increased an acceptance of nuclear ships in the ports visited.

A chronology of significant events in the NSS program is shown in Table 1-1.

The NSS completed its program of domestic and foreign port visitation in March 1965. From that time until July 1970, the ship was in commercial cargo operations. The ports visited are listed in Table 1-2.

Rev. V I

Updated Final Safety Analysis Report - (STS-004-002)

During the period of 1968 through 1970, future use options for the ship were studied and proposed.

These included the following:

Conversion to: Proposed by:

Fish Protein Concentrate Factory Ship Marine Protein Incorporated Ocean Science Ship Westinghouse Electric Corp.

Marine Nuclear Training Center Todd Shipyards Corp.

Containership including new Reactor and Steam AEIL Generators These options were ultimately rejected, and the ship was removed from service. The ship was permanently shutdown and placed in wet layup in November 1970. In Fall 1971, the nuclear plant was permanently defueled. In March 1976, the ship's primary, auxiliary, and secondary systems were drained as completely as practical.

The ship was bareboat chartered 1 to the Patriots Point Development Authority and berthed at the Patriots Point Naval and Maritime Museum, Mt. Pleasant, SC from 1981 through 1994. The ship was dry docked in Baltimore, MD for hull maintenance and preservation in 1975 and 1994. From 1994 to 2006, the ship was moored at the James River Reserve Fleet, Ft. Eustis, VA. In August 2006, the ship was moved to Colonna's Shipyard, Norfolk, VA for infrequent, required topside maintenance.

The NSS is one of the most historically significant ships in existence. It is one of the first ships to be registered as a National Historic landmark. It is registered as an International Historical Mechanical Engineering Landmark by the American Society of Mechanical Engineers. It is a registered landmark of the American Nuclear Society.

Table 1-1 Chronology of Significant Events President Eisenhower announced a proposal to build a nuclear powered merchant ship to demonstrate peaceful uses of the atom.

Congress passed P.L. 848, 84th Congress, authorizing construction of a July 30, 1956 nuclear powered merchant ship.

October 1956 Contract to Babcock & Wilcox to design and construct the reactor.

MOU established between MARAD and AEC to perform tasks related to January 22, 1957 design development, testing, initial operation and international acceptance of nuclear merchant ships.

April 1957 Contract to George Sharp to design hull.

December 10, 1957 Contract to New York Ship building to construct ship.

May 22, 1958 Keel laid.

1A bareboat charter is an arrangement where the owner gives possession of the ship to the charterer but no crew or provisions are included as part of the agreement. The charterer is responsible for all operating expenses.

Rev. V 2

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Table 1-1 Chronology of Significant Events Contract signed with States Marine Lines, to operate ship under General Agency Agreement (AEC and MARAD).

July 21, 1959 Ship launched, crew training began.

May 1960 Selection of Todd Shipyard Corporation for nuclear maintenance.

July 1960 Component testing started.

November 1960 Systems testing started.

March 1961 Public hearing.

July 24, 1961 Authorization of AEC fueling and operation for test and demonstration purposes.

September 1961 Pre-fueling testing complete.

November 11, 1961 Reactor Core installed.

December 21, 1961 Initial criticality.

1961 U.S. Coast Guard Certificate of Inspection issued (Doc. No. 287392),

Passenger Nuclear.

January 1962 10% testing complete.

March 1962 Initial sea trials at Camden, NJ and Yorktown, VA.

April 19, 1962 States Marine Lines (Operator) under General Agreement.

May 1, 1962 Delivered to and accepted by the Government.

AEC provided interim sea and port operation authorization. Amended November 19, 1962 and May 1, 1963.

Commencement of initial voyage. First voyage to commercial port August 1962 Yorktown, VA to Savannah, GA. Demonstration phase of operations began with voyage from Savannah, GA to Norfolk, VA.

February 1963 Initial voyage complete.

May 1963 Cancellation of General Agent contract with States Marine Lines, Inc.

May 1963 to Spring Labor dispute halted operations.

1964 June 1963 New staff formation.

New Contract signed with American Export Isbrandtsen Lines (AEIL) to July 23, 1963 operate ship under General Agency Agreement from 1964-1965 (New crew trained).

February 20, 1964 AEIL sea trials.

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Table 1-1 Chronology of Significant Events Demonstration phase continued (May 1964) voyage from Galveston to East Coast, U.S. ports.

1964-65 Called at U.S. and Western European ports.

May 28, 1964 Resume port operation.

Port visitation program complete and demonstration phase ended Mobile, March 1965 Alabama to Galveston, TX for annual drydocking, maintenance and inspection.

May 11, 1965 First Atomic Ship Transport, Inc. (FAST) and AEIL agreement to operate as a commercial cargo freighter.

May 18, 1965 USCG Certificate of Inspection Renewed as Cargo Nuclear.

August 1965 AEC Operating License (NS-1) issued.

Experimental commercial operations began. Bareboat chartered to First August 1965 Atomic Ship Transport, Inc. (FAST) (a subsidiary of AEIL).

September 1965 Begin scheduled commercial operation.

Commercial marine refueling at MARAD Refueling Facility, Todd Shipyards, Galveston, TX. The "fuel shuffle" was intended to extend the life of Core I. At the end of Core I lifetime, a shuffle operation was conducted to configure Core Ia. In this operation the four center elements were removed and replaced with four spare elements, two of 4.6% enrichment and the other two of 4.2%. The remaining 28 elements were rearranged to increase August to October 1968 reactivity. In general, the inside elements were moved out and the outside elements were moved in. Additionally, one control rod was replaced.

During the last voyage before the Fuel Shuffle Outage, evidence of a minor fuel failure was detected. It appeared that small amounts of fission products were released to the primary coolant whenever there was a significant change in reactor power level. Post shuffle operation indicated that the situation still existed; however, it did not limit operation or access anywhere on the ship.

The National Environmental Policy Act of 1969 requires Environmental January 1, 1970 Impact Statements for proposed actions. Past actions such as the initial NS-I operating license are exempt.

Commercial Operations ended. During experimental commercial operation August 1965 - July 1970, 32 domestic ports and 45 foreign ports were visited.

July 25, 1970 NSS berthed at Pier "E" Todd Shipyards, Galveston, TX and Reactor Shutdown July 29, 1970.

Final Voyage to Pier "E" Todd Shipyards, Galveston, TX from annual November 6-8, 1970 drydocking 10/30 through 11/06 at Todd Shipyards, New Orleans, LA for maintenance per Operations Report, FAST-21 April 26 - November 9, 1970.

November 8, 1970 Final Reactor shutdown at 5:50 PM and established Cold Iron condition at Rev. V 4

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Table 1-1 Chronology of Significant Events Todd Shipyard, Galveston, TX per FAST-21.

November 9, 1970 License NS-I transferred from FAST to MARAD (License Amendment 7).

April 13, 1971 AEC approved Technical Specification Change 11 to defuel NSS.

Ship at MARAD Refueling Facility, Pier "E" Todd Shipyards, Galveston, TX, for preparation of vessel for lay-up. Started removing fuel August 23, 1971. The last fuel element was removed September 29. Control rod drive (CRD) system disabled and sealed plates installed on 21 CRD head nozzles to separate drives from the control rods. 36 Fuel element flow nozzles were loaded into fuel channels in the reactor vessel. One control rod (cut into two Fall 1971 six feet long pieces and one twelve inch piece) was placed in reactor vessel.

This is the control rod removed during the 1968 fuel shuffle. RCP motors and impellers were removed. The RCP volutes were blank flanged. Reactor, secondary and saltwater system filled with water to establish wet lay-up conditions. The reactor vessel head was reinstalled with 6 of the normal 48 studs in place. The containment vessel, port and starboard charge pump room and lower reactor compartment bilges were decontaminated, washed, vacuumed and painted. All cargo holds were decontaminated.

October 24, 1971 U.S. Coast Guard Certificate of Inspection expired and was not renewed.

Towed to Savannah, GA and following a welcoming celebration, additional January 3-10, 1972 non nuclear layup work was performed (4/24/1972 Operations Report) [Not open to Public and primary system and steam generators (boilers) maintained in wet lay-up (MARAD letter Nov 19, 1971)].

AEC stopped review of Core II and noted "If reactivation of N. S. Savannah April 4, 1972 should become planned, an appropriate scope of review would need to be established in light of... the regulatory safety and environmental requirements at that time."

All 36 Core I Spent fuel elements (32 plus the four replaced during the fuel shuffle) were transferred by AEC for reprocessing in SC.

License Amendment (Technical Specification Change 13) recognized the ship was in a "Fuel Removed Condition." Note that even though the final shutdown was in November 1970 and the permanent defueling was January 29, 1973 completed in fall 1971, these actions were not considered to be permanent cessation of operations until the decision was finally made to not load Core II.

Therefore, the fall 1971 defueling became the de facto permanent defueling and cessation of operations with the issuance of Technical Specification Change 13.

1972-75 Legislation was introduced to authorize the Secretary of Commerce to transfer ship to the City of Savannah, GA.

MARAD memo "Supplemental Memo on Working Document.

January 31, 1975 EnvironmentalImpact Analysis NuclearMerchant Ship Program" -MARAD recommended moving Core 11from storage at Defense Atomic Support Rev. V 5

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Table 1-1 Chronology of Significant Events Agency, Killeen, TX to Nuclear Weapons and Storage Facility, Manzano Air Force Base, Albuquerque, NM for indefinite storage. This action would be taken via an Interagency Agreement by Energy Research and Development Administration (ERDA) (formerly AEC). Per NSS Defueling Log September 9, 1971, (MMA File 61-8), Core II consists of 36 new fuel elements.

Attachment I to memo for Assistant Administrator for Commercial April 3, 1975 Development states "In response to M-700 memo of December 23, 1974,

5. Nuclear Fuel (Core 11) it is understood this fuel will be returned to AEC

[ERDA]."

July 28, 1975 City of Savannah, GA requested removal of the ship and the ship left Savannah, GA for Baltimore, MD drydocking.

Moved from Baltimore, MD Drydock to Detyens Shipyard, Inc. Charleston, SC (topside work).

Moved from Detyens Shipyard, Charleston, SC to N. Charleston Army August 29, 1975 Depot, SC pending the outcome of legislation before the Congress to permit use of the NSS as an addition to the Patriots Point Naval and Maritime Museum, Mt. Pleasant, SC.

While at N. Charleston Army Depot, SC, the primary, auxiliary and secondary systems were drained by Todd Research and Technical Division as completely as practical.

Per the September 14, 1976, Annual Operations Report, water from the reactor vessel shield water tank and secondary side of the steam generators was transferred to a double bottom tank below the reactor compartment.

Demineralizer resins (28 Ci) and 12,200 gallons of liquid (Primary Coolant March 1976 and liquid waste) were properly disposed. Three dewatered primary purification ion exchangers (demineralizers) were removed and disposed.

Radiation levels in the containment vessel did not increase appreciably after the water was removed.

Per March 9, 1976 MARAD memo "Status Report on the Transfer of N.S.

Savannah to [James River Reserve Fleet] JFFR," the water from the secondary system was analyzed and does not have to be disposed of as nuclear waste.

Possession Only License issued (License Amendment 8) and recognized the May 19, 1976 ship was in a state of protective storage. Total estimated residual activity 1.09E+5 Ci.

April 19802,18) Main reduction bull gear was removed (NRC Inspection Report 81-01, March 2, 1981).

Public law 96-331 was enacted to authorize the NSS to be sent to Patriots August 28, 1980 Point, Mt. Pleasant, SC for use as a Museum and other public purposes.

September 24, 1980 Bureau Radiological Health, SC Department of Health & Environmental Rev. V 6

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Table 1-1 Chronology of Significant Events Control certifies all ten remaining radioactive sources have been transferred from the ship to them.

Patriots Point Development Authority (PPDA) became a co-licensee and the August 14, 1981 ship was bareboat chartered for public display at the Patriots Point Naval and Maritime Museum, Mt. Pleasant, SC from 1981 through 1994.

The American Society of Mechanical Engineers designated the ship as the Fourteenth International Historic Mechanical Engineering Landmark.

July 15, 1986 The NS-I License was renewed for ten years by License Amendment 11.

September 1989 Hurricane Hugo hits the Charleston, SC area.

The National Park Service designated the ship as a National Historic July 17, 1991Ladrk Landmark.

The Board of Directors for the American Nuclear Society approved the ship for a Nuclear Historic Landmark Award.

Shortly after the second bareboat charter renewal in 14 July 1991, a hull leak July 1991 developed. After a series of surveys and inspections, funding was requested from Congress in 1993 to drydock the ship to correct deterioration of the underwater hull.

December 1993 Public Tours of the ship were no longer allowed.

NSS left Patriots Point on the spring flood tide to minimize silting effects from Hurricane Hugo of September 1989.

License Amendment 12 removed PPDA as co-licensee. After routine drydocking and hull maintenance in Drydock # 4, Sparrows Point, Baltimore, MD, the ship was placed in protective storage at the James River Reserve June 29, 1994 Fleet, Ft. Eustis, VA. Numerous minor modifications including installation of a DH system were completed. Modifications associated with the DH system (i.e., vent ducting and cabling) generally obstructs the doorways, passageways and ladders throughout ship as is typical in the marine industry for deactivated ships.

License Renewal was no longer required by the revised 10 CFR 50.51, Continuation of license.

(b) Each license for a facility that has permanently ceased operations, July 29, 1996 continues in effect beyond the expiration date to authorize ownership and possession of the production or utilization facility, until the Commission notifies the licensee in writing that the license is terminated.

NSS was moved to Colonna's Shipyard, Norfolk, VA for infrequent, required topside maintenance.

December 11, 2006 Post Shutdown Decommissioning Activities Report, Revision 0 was submitted but withdrawn on January 27, 2007.

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Table 1-1 Chronology of Significant Events January 30, 2007 NSS moved to Pier 23 Newport News, VA for temporary layberthing.

License Amendment 13 was issued. It included six administrative changes.

January 31, 2007 The most significant change is following 30 day notification to NRC, the ship can be located at any appropriate U.S. domestic location with a MARAD approved Port Operating Plan.

NSS was moved to BAE Systems Norfolk Ship Repair facility in South Norfolk, Virginia in preparation for drydocking. On January 19, 2008, the NSS was placed onto the Titan Drydock and remained there until March 3 for May 31, 2007 hull maintenance and exterior surface cleaning and preservation. Hull inspections were conducted and identified no new degradation. These inspections included verifying the integrity of the underwater hull repairs made during the 1994 drydocking availability License Amendment No. 14 was issued. It revised the Technical Specifications to match the current ship status and ship's programs, policies April 3, 2008 and procedures. The function of Review and Audit Committee was revised commensurate with the Decommissioning Quality Assurance Plan. The committee was renamed the Safety Review Committee.

NSS was moved to Pier 13, Canton Marine Terminal at 460 1Newgate Ave.,

May 8, 2008 Baltimore, MD.

December 11, 2008 Post Shutdown Decommisýsioning Activities Report, Revision I Submitted.

1.2 PrincipalSafety Featuresof the Ship 1.2.1 The Hull This text is retained from Revision III for historical context. In particular, items (1), (3) and (4) from the list below describe structural features installed on the N.S. Savannah to enable it to resist the effects of damage. Item (2) describes the hydrodynamic standard applied to allow the ship to passively resist the effects of damage and remain afloat; this condition is referred to as "damaged stability." During the period of museum operations at PPDA, most of the transverse watertight bulkheads listed in (1) were penetrated at various locations to permit convenient personnel access and service connections between the compartments. Most of these penetrations were closed prior to the ship's removal from PPDA in 1994. Current operations only require the NSS to meet "intact stability" requirements, where intact stability refers to the capability of the ship to remain stable when not damaged. Based on stability calculations prepared in 1994 and the current (2008) baseline structural condition assessment, the NSS is presumed capable of meeting a "one compartment" damaged stability standard For a ship of its generation, the hull design of the N.S. Savannah was that of a conventional vessel with trim and modem lines, designed in accordance with the latest practice in naval architecture and hydrodynamics.

The principal safety features in the hull design may be summarized as follows:

1. Thehull is divided by 10 watertight bulkheads terminating at the B-deck level.

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2. The ship will remain afloat if any two of the nine compartment are flooded within the meaning of U.S. Coast Guard (USCG) stability requirements.

3. Conventional all-welded steel construction is employed with full-continuity of structural members, except for a few longitudinal riveted seams serving as crack arresters in accordance with contemporary practices.

4. A 5-foot deep watertight double-bottom extends under all nine compartments but not under fore peak or after peak ballast tanks.

5. Special care was taken in the selection of materials to minimize galvanic corrosion.

6. The reactor compartment is reinforced outboard of the containment vessel with additional longitudinal and athwartship bulkheads, in addition to a laminated redwood and steel collision mat on each side, to minimize penetration in the event of a collision amidships.

7. The design details and construction of all components and systems was in accordance with the contemporary regulations of the Safety of Life at Sea Convention (SOLAS), MARAD, USCG, the U.S. Public Health Service, the American Bureau of Shipping (ABS) and other regulatory bodies.

1.2.2 The Power Plant From the beginning of the project, analyses of hypothetical casualties and topical reports covering every phase of the design were prepared and reviewed by the regulatory agencies as well as Oak Ridge National Laboratory (ORNL). In addition, a comprehensive analysis of reactor safeguards was prepared for review by the AEC and for the critical examination of the AEC's Advisory Committee on Reactor Safeguards (ACRS).

1.2.2.1 General In the design and construction of the power plant, a special philosophy of safety and reliability was adopted at the outset of the project. Both preparatory and concurrent research and development work and special operational testing were undertaken in many areas, including a series of critical experiments on various configurations and designs for the reactor core, full scale testing of a prototype control rod drive mechanism and system under simulated reactor conditions, irradiation tests of fuel elements and pellets, containment vessel leak rate tests and filter efficiency tests.

1.2.2.2 Particular Safety Features The NSS has four remaining safety features in the power plant design that merit special mention:

1. Reactor System Containment - The reactor and primary coolant system are located inside a containment vessel. During the operating period, any failure of the primary system would have been contained by the containment vessel system. The CV system, consisting of a USCG approved pressure vessel and associated isolation valves, was designed and hydrostatically tested to withstand a pressure of 173 psig. During the operating period, the system was routinely pneumatically tested at 60 psig and demonstrated to have a leak rate of less than 1.2% of the contained volume per day. No integrated leak rate testing has been performed since the final shutdown in November 1970.

2. Containment Vessel Isolation - During the operating period, the reactor system was operated with the CV hermetically sealed at all times. The penetrating pipe lines were Rev. V 9

Updated Final Safety Analysis Report - (STS-004-002) isolated by automatic valves if the pressure inside the containment vessel rose above 5 psig for any reason whatsoever. The steel containment vessel, moreover, was sufficiently thick that no danger of penetration existed from any high speed missiles within the containment vessel, such as a thermocouple well from the primary coolant system. No individual penetrations have been leak rate tested since the final shutdown in November 1970.

3. Radioactive Waste Control - During the operating period, the power plant operated with a minimum of radioactive waste disposal. Gaseous and liquid wastes were discharged in accordance with the "Standards for Protection Against Radiation" (10 CFR Part 20) and the Technical Specifications. Any solid wastes are kept aboard the ship and periodically discharged to a licensed waste disposal facility.

4. Shielding - During the operating period, the reactor vessel was surrounded laterally by 33 inches of water and I to 4 inches of lead shielding so that maintenance work was possible shortly after shutdown. In 1976, the reactor vessel shield water tank was drained. The secondary shielding around the containment vessel reduced the surface dose level at full reactor power to no more than 100 mR per week.

1.3 Ship and System Status - 1976 to Present 1.3.1 Power Plant Status as reported in April 1976 In support of the request for a possession only license in 1976, MARAD provided a "current" (i.e.,

contemporary) status of the ship. The April 1976 status of the NSS reactor plant following protective storage modifications to its operating condition was noted as follows:

Reactor Vessel - All 32 Core Ia fuel elements were removed from the reactor vessel to the spent fuel pool at the MARAD Refueling Facility, Todd Shipyards, Galveston, TX [License SNM-1015-YGP (Docket 70-1062)]. All reactor internal components are in place, i.e., core basket, upper grid plate, upper flow baffle. Thirty-six upper flow transition pieces and one irradiated control rod (cut in three pieces) have been loaded into six fuel element channels in the core basket. Thirty-two upper flow transition pieces were from Core I and four were from Core Ia (i.e., the 1968 fuel shuffle). The additional control rod was the one replaced during the 1968 fuel shuffle. The reactor head is in place with six of the original 48 reactor head hold down studs tensioned.

2. Spent Fuel - Per September 21, 1973, Operations Report, thirty six spent fuel elements (Core I and la) were shipped from Galveston, TX to USDOE - Savannah River Plant, Aiken, SC in nine shipments from October 4 through December 21, 1972 via a lowboy trailer using the Piqua/Elk River Shipping cask.

3. Fission Chambers - Per September 21, 1973, Operations Report, five fission Chambers were removed and shipped to Todd Shipyards, Galveston, TX in May 1973. On June 28, 1973, these five and an additional three stored on the NSV Atomic Servant (the nuclear services barge for NSS) were shipped to Chem-Nuclear Systems, Inc., Barnwell, SC for disposition as radioactive material.

4. Pressurizer - One relief valve PR-1V was removed. The nozzle flange was fitted with an absolute filter vent to allow for pressure equalization. Circuit breakers for the heaters have been opened at the switchboard.

5. Control Rod Drive System -

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a. Electrical - The breakers for the rod drives have been opened and tagged in the Control Rod Drive Electrical Cubicle and the Control Room. The cables between the junction boxes on the Control Rod Drive (CRD) structure and the junction boxes located around the periphery of the cupola have been disconnected, coiled and secured on the CRD structure in the containment vessel. The support wireways have been removed, packaged and stored in Hold No. 4. In addition, the twenty one buffer seal flow meter cables and the vibration monitor cables have been disconnected.

b. Mechanical - All hydraulic oil has been drained and the piping disconnected between the cupola and the CRD structure. Buffer seal system valves have been closed and tagged out.

The inlet and outlet header spool pieces between the cupola and structure were removed.

The buffer seal system has been drained.

The 21 control rods were disconnected from their respective CRD extension shafts and remain in their fully inserted position between the empty fuel location channels. The CRD extension shafts have been disconnected from the lead screw and are in a fully inserted position from the reactor head nozzles to the top of the control rods. The CRD lead screws were withdrawn and pinned in place. Steel caps are placed onto the 21 reactor head nozzles to completely seal the nozzle from the containment vessel atmosphere and mechanically separate the CRD mechanism from the control rod.

6. Primary Coolant System - The system has beendrained as completely as practical. All loop isolation inlet and outlet valves are open and back seated.

Primary Pump motors and impellers were removed. Blank flanges are installed over the volute openings. The grating, piping, valves, ducting, etc. that were disconnected while removing the pump motors and impellers are left in the removed condition within the containment vessel.

7. Secondary System - Both steam generators and piping were drained of water as completely as practical. The water on the secondary side of the steam generators was transferred to the double bottom tank below the reactor compartment. The isolation valves outside of the containment vessel are closed. There have been no modifications to the secondary system.

8. Auxiliary Systems - All nuclear related auxiliary systems have been drained as completely as Table 1-2 Tank Designation practical. SL-TI - Buffer Seal Surge Tank The auxiliary system tanks listed in Table 1-2 have been drained of liquids. Some still retain non- PD-T1 - Lab Waste Tank removable surface contamination. PD-T2 - Makeup Storage Tank The three primary purification system ion- PD-T3 - Makeup Storage Tank exchangers (PP-El, PP-E2 and PP-E3) and their resins (28 Ci) have been physically removed from PD-T4 - Containment Drain Tank the ship and their lines capped. PD-T5 - Double Bottom Tank Water from the primary (neutron) shield water tank PD-T6 - Double Bottom Tank was transferred to the double bottom tank below the reactor compartment. Primary (Neutron) Shield Tank

9. Containment Vessel - The forward and aft manway PR-TI - Effluent Condensing Tank shield plugs were put in place with security seals installed. Containment vessel (CV) entries were expected to be made through the aft airlock.

The lifting chains and hooks were removed from the chain hoists used to handle the shield plugs Rev. V I1I

Updated Final Safety Analysis Report - (STS-004-002) to minimize the possibility of unauthorized containment vessel entry. The CV bilge was wiped clean. The inerting piping in the CV cupola head was disconnected outside and valves closed so that the integrity of the vessel was maintained.

10. Control Areas - The following areas of the ship are controlled by locking because they contain radioactive systems or components:

Secured boundaries were established to prevent unauthorized access into Radiation Control Areas (RCAs). The secured entrances included the following spaces:

a. The reactor compartment surrounding the containment vessel that originally formed a secondary containment around the pressure containment;

b. The Port and Starboard Stabilizer rooms;

c. The Forward Control which includes on C-Deck the Cold Water Chemistry Lab (Port) and Radiation Sampling Room plus on D-deck the Gas Adsorption Equipment Room (port) and Radiation Sampling Room (stbd);

d. The Port and Starboard Charge Pump rooms; and,

e. The Hot Chemistry Laboratory, in the Control Room Area.

When not in use, the entries to all of the above spaces are kept locked and are free of surface contamination (less than 450 pCi/100 cm 2) except for the interior of the containment vessel and lower reactor compartment. The containment vessel, port and starboard buffer seal charge pump room and lower reactor compartment bilges were decontaminated, washed, vacuumed and painted in Fall 1971 as part of defueling.

A complete survey was made on the ship after removal of the primary system water and demineralizers (ion-exchangers). Radiation levels in the containment vessel did not increase appreciably after the water was removed.

All loose radioactive material (components and solid waste) was removed from the ship. At present, there is no radioactive or contaminated material outside of control areas.

The calculated residual radioactivity remaining in the plant as of April 1976 as listed below:

Source Activity Activated components within the Reactor Vessel 108486 Ci Corrosion and Fission products plated out throughout the system 16.6 Ci Total Estimated Residual Activity 108513 Ci

11. All of the hull openings have been welded closed. The only hull opening with the potential for a leakage path is the main propeller shaft bearing in the stern tube. (2009 -Note that as part of the 2008 drydocking work, the stern tube opening was permanently sealedand capped after removal of the tail shaft andpropeller.)

1.3.2 Status of N.S. Savannah from April 1976 - April 2007 During the intervening years, the ship has remained in a state of protective storage.

In 1981, the NSS was bareboat chartered to the Patriots Point Development Authority (PPDA) of the State of South Carolina, for public display at the Patriots Point Naval and Maritime Museum, near Rev. V 12

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Charleston, SC. During that period the PPDA was designated a "co-licensee" for the reactor and exercised custody of the ship - but ownership remained with MARAD.

Shortly after the second bareboat charter renewal in 1991, a hull leak developed. After a series of surveys and inspections, it was concluded in 1993 that deterioration of the underwater hull body necessitated a drydocking, for which funding was requested from Congress. Unrelated to the underwater hull issue was internal deterioration from excessive water intrusion from the upper decks.

The water intrusion condition began with Hurricane Hugo in 1989 and gradually worsened over time. In light of these issues and the effects they had on use of the ship, PPDA informally notified the Agency that it would exercise its right to terminate the charter effective with the ship's removal for drydocking.

Museum operations ended in early 1994, when industrial topside and internal repairs commenced in preparation for the drydocking and subsequent transfer of the NSS to a reserve fleet. The NSS was removed from the museum site on May 24, 1994 and in June 1994 was drydocked in Baltimore, MD for hull maintenance and repair. Extensive repairs to deteriorated underwater hull plating were completed, and in July 1994 the NSS was placed in the Agency's James River Reserve Fleet (JRRF) at Fort Eustis (Newport News), VA. License Amendment 12 was issued in July 1994, and among other things removed the PPDA as a co-licensee, and established the vessel's location as the JRRF.

The NSS remained in protective storage there until August 15, 2006.

To prepare the ship for layup at JRRF, a number of conventional preservation measures were put into place. These included fitting a Dehumidification (DH) system to preserve the ship's interior, renovations to the Impressed Current hull Cathodic Protection (ICCP) system, and overhaul of the ship's existing fire, flooding and intrusion alarm systems. These modifications included long runs of portable ductwork and electrical cabling which generally obstructed the doorways, passageways and ladders throughout ship. Externally, deck drains were blanked off to reduce water intrusion into the ship's interior. No funds were available to correct the interior water damage effects; however, the DH did arrest any further degradation.

In preparation for, and during the lay-up period at the JRRF, no changes were made to radiological surveillance or monitoring plans and no substantive radiological work was performed. Radiological work c*onsisted exclusively of passive surveillance and monitoring in accordance with the circa 1981 surveillance plan. Only minimal modification, maintenance and testing of the alarm and monitoring systems required by Technical Specifications ("TS") were performed.

In March to May 2005, a radiological and environmental characterization scoping survey of the NSS was conducted. Based on initial results of the survey, intrusive sampling and analysis was conducted on the reactor pressure vessel and related components. This sampling found that those components can be disposed of as Class A radioactive waste material at qualifying radioactive waste disposal facilities.

In May 2005, MARAD completed an internal reorganization in which resulted in reestablishing the Savannah Technical Staff (STS). The STS is now the organizational unit within MARAD that is responsible for all NSS program activities.

The Maritime Administration's Senior Technical Advisor (STA) has overall responsibility for administration of the NS-1 license and will provide guidance and oversight to the Manager-N.S.

Savannah Programs (MSP). The MSP is responsible for the execution of all program activities including routine radiation surveillance, ship custody decommissioning and historic preservation.

MARAD may assign a single incumbent to both positions. This individual is assisted by a Decommissioning Program Manager, a Documentation Manager and a Facility Site Manager.

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A significant portion of the decommissioning project work is intended to be performed by contractors. MARAD personnel, reporting directly to the MSP, will be responsible for assigned contractors and providing the contractor direction, coordination and interfacing communications.

In May 2006, further organizational enhancements were made. Consultants were contracted to fill the roles of Radiation Safety Officer, Quality Assurance Manager, Risk Manager and Licensing and Compliance Manager. The QA Manager will provide independent verification of contractors and the contractor's QA Manual which will be approved by the NSS Project. Individuals from the Argonne National Laboratory and the Volpe National Transportation Systems Center were contracted to provide additional expertise in Decommissioning Environmental and Maritime issues.

On August 15, 2006, the NSS was moved to Colonna's Shipyard, Norfolk, VA for infrequent, required topside maintenance. On January 30, 2007, the NSS was moved to Pier 23, Newport News, VA to prepare for drydocking.

1.3.3 Radiological and Environmental Characterization Scoping Survey In September 2005, a radiological and environmental characterization scoping survey of the NSS was conducted. This characterization task was intended to provide a basis for estimating the cost of decommissioning. It is documented in NSS Radiological and Non-Radiological Spaces Characterization Survey Report and further described in Chapter 13.

The principle findings of the characterization scoping survey include the following:

The CV entries were the first in over thirty years. The observed conditions generally supported the reported conditions of 1976. However, discrepancies were noted such as presence of hydraulic oils, migration of primary system water, significant quantities of trash/debris and superficial surface corrosion of components and lighting fixtures.

" Absence of fission products (other than trace quantities of cesium-137), uranium and its daughter isotopes, as well as transuranics (e.g., plutonium), indicates there were no significant fuel failures;

Crud contribution to total curie content is minimal;

No contamination was found in the non-radiological spaces;

Minimal contamination was found in radiological spaces;

Overall dose rates were much lower than expected;

" Sites previously identified as radiologically contaminated were found uncontaminated; and,

Containment vessel systems, structures and components were exceptionally radiologically clean.

The scoping survey concluded that the NSS is in very good condition from a radiological perspective to support decommissioning. Note that during the assessment approximately 1500 gallons were found in the primary system. This volume is approximately 15% of the normal operating water volume of the system (approximately 10,000 gallons).

1.3.4 Current Status of N.S. Savannah April 2007 to present The ship is in a state of protective storage.

On May 31, 2007, the NSS was moved to BAE Systems Norfolk Ship Repair facility in South Norfolk, Virginia in preparation for drydocking.

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In 2007, MARAD continued its on-going efforts of improving the technical capabilities of the Savannah Technical Staff (STS) and filled the following positions with nuclear and/or maritime experienced staff:

" Nuclear Advisor;

" A Partial Ship's Crew - typically comprised of the following:

First Engineer; a Electrician; and,

Two General Vessel Assistants.

The result of this increase in staffing allows the ship to be routinely manned during a normal work week. This increase in staffing continues to foster improvement in the physical and administrative aspects of the NSS programs.

On January 19, 2008, the NSS was placed onto the Titan Drydock at BAE Systems Norfolk Ship Repair. The NSS remained there until March 3. In addition to the routine need to inspect and repair the ship's underwater hull, MARAD used this maintenance availability to perform a comprehensive assessment of the physical condition of the ship's primary structure and exterior envelope.

As part of the assessment, MARAD entered into an agreement with the American Bureau of Shipping (ABS) to return the ship to hull classification. Once classed, the ABS will provide independent monitoring and review of the ship's structural condition.

At the previous drydocking availability in 1994, extensive repairs were made to the underwater hull by clad-welding deep pits; renewal of several hundred linear feet of welded butts and seams; and installation of steel doubler plates over four areas of generalized erosion and pitting at the "wind and water" interface.

These repairs were subjected to intense survey and inspection by ABS and STS during the 2008 drydocking. All of the repairs were found to be in satisfactory condition. Three of the doubler plates were removed, and several of the underlying locations were inserted with new steel plates. The fourth doubler plate is located adjacent to the Cold Chemistry Laboratory, and was thus not removed because of the need to dismantle the laboratory equipment (with requisite radiological controls) to provide access for the required hot work in that area.

No record of ballast tank or other innerbottom tank inspections was found for the 1994 or the preceding 1975 drydocking availabilities. Consequently, planning for the 2008 drydocking availability included inspection and survey of all innerbottom and ballast tanks. All tank surveys and inspections were completed, except for ballast and innerbottom tanks within the Reactor Compartment and the number 4 centerline ballast tank2 . Ultrasonic thickness gaugings were taken of thecargo hold tank tops (a principal structural member of the ship), and the ship's exterior hull structure. Additional gaugings were taken adjacent to each cargo hatch opening in the main deck; on the cargo hatch covers themselves, and at selected areas of structural concern and interest. The results were satisfactory.

2The No. 4 centerline ballast tank is filled with water containing chemical compounds of indeterminate origin. The tank will be inspected at a future date after the water is processed and discharged. Structural inspections were completed of the exterior of the tank boundaries and found satisfactory.

Rev. V 15

Updated Final Safety Analysis Report - (STS-004-002)

The entire exterior surface of the ship was prepared by waterjetting. The underwater hull, freeboard and main deck areas were prepared to bare steel, with all previous and original coating materials removed. A new high performance epoxy coating system was applied, with high gloss acrylic urethane topcoat. The underwater hull received one coat of conventional anti-fouling coating. Other areas of the topside exterior were prepared to a slightly lesser standard, with some original coating materials left intact. All previously existing hydraulic oils and similar materials were removed from non-functioning deck equipment, to minimize the risk of discharge of such materials into the environment.

The exterior envelope of the ship forms the perimeter boundary for licensed activities. Maintenance of the envelope ensures that the nuclear facilities housed inside the ship are protected. Based on the review of available records, it is surmised that the 2008 drydocking was the first comprehensive examination of the ship's structural fabric since 1967, when the ship underwent "Special Survey No.

I" for the ABS.

Major drydocking work completed:

Underwater hull cleaning, surface preparation and coating;

" Inspection and repair of 1994 hull repairs;

" Freeboard preservation from deep loadline to top of the rails;

" Topside and superstructure preservation including weather exposed decks;

Audio gauge exterior hull, decks, superstructure, tank tops and tanks;

" Cathodic protection system; service and repair;

" Removed ship's propeller and stored it on No. 7 cargo hold hatch cover;

" Permanently sealed and capped the stem tube opening after removal of the propeller;

" Examined, tested and repaired (as required) all hull blanks;

" Examined and repaired as required bilge keel and bilge strake rivets;

" Performed miscellaneous steel renewals, as needed, to side shell, bottom and main deck plating;

" Removed ceramic tiles and magnesite underlayment from the Promenade Deck to permit inspection, repair and coating of underlying steel decks;

Internally inspected double bottom ballast tanks, forward and after peak tanks, No. I Deep Tank, and various voids and pipe tunnels;

Replaced the weathertight seals on main deck hatch covers at cargo holds 1, 2, 3, 4 and 7.

In summary, the structural condition of the ship was found to be very good. In particular, the innerbottom structure was found to have little-to-no structural degradation, and the original tank coatings were found intact and serviceable. All major hull strength components were inspected, repaired if necessary, and coated for long-term preservation. The thorough inspections and surveys of the ship's structural condition were documented to provide a baseline for the balance of the protective storage period.

On May 8, 2008, the NSS was moved to Pier 13, Canton Marine Terminal at 4601Newgate Ave.,

Baltimore, MD. Maintenance activities continued once the ship was moved to its Baltimore layberth. From that point onward, the maintenance strategy focused on routine preventative maintenance, and restoration of ship systems and equipment necessary for husbanding the ship, and long-term retention. In addition, discrete activities were performed to provide for more secure access control, security surveillance, and protection of shipboard personnel. Infrastructure improvements required to employ shipboard habitability areas for routine administrative purposes such as offices and meeting rooms are ongoing.

Rev. V 16

Updated Final Safety Analysis Report - (STS-004-002)

Completed and Ongoing Maintenance Activities at Layberth:

Modification of port-side Passenger Sideport for secure daytime access;

Modification of Stateroom A- 18 to Ship Security Office (adjacent to sideport);

Installed a new security gate key fob reader pad system at the sideport entrance;

" Reactivated the General Alarm to be audible throughout the ship;

" Cleaned and painted numerous interior areas of the ship;

Gutted three adjoining crew staterooms on B Deck and converted area to Radiological Control Station;

Gutted six adjoining crew staterooms on C Deck and converted area to a Records Vault;

HVAC installation on "B" Deck for offices / conference room; and "C" Deck Technical Library and Baggage Room (document storage);

" Electrical distribution system surveys and inspections; restoration of interior and exterior lighting; and,

Repairs to aft mooring capstans and controllers.

1.4 Symbols for the Reactor PlantAuxiliary Systems In the text, the Chapter 9 Reactor Plant Auxiliary Systems are referred to by the symbols listed below:

System Symbol Containment Cooling CC Control Rod Drive CRD Intermediate Cooling Water CW Data Acquisition DA Emergency Cooling DX Hydrogen Addition HA Nuclear Instrumentation NI Equipment Drain and Waste Collection PD Primary Pressurizing PE Primary Loop Purification PP Primary Relief PR Radiation Monitoring RM Reactor Space Ventilation RSV Sampling SA Shutdown Circulation SC Buffer Seal SL Soluble Poison SP Waste Dilution and Disposal WD Gaseous Waste Disposal and Collection WL Rev. V 17

Updated Final Safety Analysis Report - (STS-004-002) 2 SHIP DESCRIPTION 2.1 Featuresand Characteristics Prior to permanent defueling in Fall 1971, the N.S. Savannah (NSS) operated as a single-screw, all-steel, nine-compartment, cargo ship, with raked bow and modified cruiser stem. The ship has three complete decks and three upper decks in the superstructure. The ship's principal characteristics are shown in Table 2-1.

Table 2-1 Principal Characteristics of the N.S. Savannah Dimensions Feet Length overall 595.5 Length fore-post to rudder-post 545 Beam, molded 78 Height, baseline to weather deck (A-deck) 50 Height, baseline to top of house 85 Draft, design 29 Draft, light ship (approximate) 18.33 Displacement and Tonnage in Long Tons Net carrying capacity, total deadweight tons 9,656 Full load displacement, tons 21,990 Light ship, tons 12,334 2.2 Compartmentation,Decks and Space Arrangement 2.2.1 Compartmentation Following reconstruction of bulkhead openings created when the ship was on public display at the Patriots Point Naval and Maritime Museum, the integrity of the ship's transverse watertight bulkheads was not confirmed (see Section 1.2.1). Therefore, the ship's original design capacity to resist damage and remain afloat with any two compartments flooded cannot be verified. The NSS meets all applicable USCG intact stability criteria, which is the minimum standard required for ship movements and layberthing.

The following text is retained for historical purposes.

The hull is subdivided by 10 watertight transverse bulkheads extending to the freeboard deck (B-deck). The ship is designed to a two-compartment standard of subdivision (i.e., the ship will remain afloat with any two compartments totally flooded at the vessel's full load draft of 29 feet 6 inches with the margin line 3 inches below B-deck), which meets the requirements of the USCG. Tonnage openings have been closed to convert the NSS from a sheltered deck to a full scantling vessel resulting in additional damaged stability.

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Updated Final Safety Analysis Report - (STS-004-002)

Beginning at the bow, the compartments and spaces between transverse bulkheads are as shown in Table 2-2.

Table 2-2 Compartment Utilization Space Length (feet) Assignment Fore Peak 28 Tank and storerooms Hold 1 50 General cargo and deep tank Hold 2 55 General cargo Hold 3 55 General cargo Hold 4 55 General cargo Reactor space 60 Containment vessel, reactor compartment, storerooms and stabilizers Engine room 55 Engine room and control room Hold 5 50 Storerooms Hold 6 57.5 General cargo Hold 7 47.5 General cargo Aft peak 32 Tank, steering gear and storerooms The hull is built on a transverse framing system except for the inner bottom, which is a combination of transverse and longitudinal framing specially stiffened below the reactor compartment. The inner bottom is also subdivided into tanks for carrying liquid ballast, diesel oil, steam generator (boiler) feedwater, fresh water and low-level radioactive liquid wastes.

2.2.2 Deck Arrangements The ship has three complete decks (A, B and C) with platform decks at various levels. A promenade deck extends over the A-deck for about one-third of the ship's length.

The NSS's streamlined superstructure is set sufficiently aft to permit access to the reactor system*

containment vessel through the forward promenade deck for refueling. The forward weather deck accommodates the hatches for cargo holds 1, 2, 3 and 4. Each hold was served by two cargo booms.

Holds 1 and 2 were served by 8-ton cargo booms and holds 3 and 4 by 10-ton cargo booms. The eight booms were operated in groups of four from two A-frame supporting structures with outriggers.

Aft of the superstructure a third cargo gear A-frame accommodates four 10-ton booms that served holds 6 and 7. Cargo hatch covers are set in coamings on A-deck and were hydraulically operated from local stations at each hatch.

All of the cargo handling equipment is deactivated, disabled and performs no active function.

The uppermost or navigation bridge deck served a dual purpose. The pilothouse is located on the forward end and the radio room is slightly aft to the starboard. The chart room is on the port side outboard of the gyrocompass room. The balance of the navigation bridge deck provided berthing space for the radio operator and is the location of two fan rooms, a battery room and the emergency generator room.

Rev. V 19

Updated Final Safety Analysis Report - (STS-004-002)

The NSS's pilothouse and navigation bridge were laid out to obtain the maximum visibility through the widest possible angle (almost 180 degrees). The pilothouse was equipped with the latest available navigation and communication equipment for a ship of its generation. The navigation control console housing all normal pilothouse instrumentation was situated near the forward end of the house and on the center-line. The magnetic compass was of the reflecting type and was the first of this type to be manufactured in the United States. The pilothouse also houses the control console for the anti-roll stabilizers which are installed on and above the inner bottom tank top level amidships. All of the navigation equipment is deactivated, disabled and performs no active function.

The boat deck, just below the bridge, is almost entirely occupied with officers' accommodations. An officers' lounge located on the after end of the deck afforded observation on either side of the ship.

Various public rooms and other spaces are located on the promenade deck. The forward portion is a weather deck containing the main hatch over the reactor compartment. Between the sloped front of the deckhouse and the main lounge, there is an enclosed passageway extending outboard athwartship from the centerline. This passageway also extends along both sides of the ship outboard of the inner house and is called the promenade.

2.2.3 Space Arrangement As seen in Figure 2-1, the ship's power plant is located in the two amidships compartments (reactor space and engine room) which lie between frames 99 and 148.

Components of the reactor system are located in the reactor space. Most of the components are inside the containment vessel. The engine room contains all of the propulsion system equipment by which high pressure steam was converted to power. The engine room also contained the main and auxiliary electrical generating systems and most of the other plant auxiliary systems.

The main control room for the entire power plant is located on the 20-foot flat at the after end of the engine room. All operations for starting and controlling the power plant were executed from this room. Almost all measurements, conditions and events in the reactor and propulsion systems were monitored and logged in this room and all power plant maintenance operations were supervised from it.

2.3 Navigation andSteering Equipment 2.3.1 Navigation Equipment All of the navigation equipment is deactivated, disabled and performs no active function.

2.3.2 Steering Gear The steering gear is mechanically locked to keep the rudder centered during towing operations. It performs no active function.

2.4 FireDetection and Control System A limited number of portable fire extinguishers and fire axes are located on the ship. Fire detectors alarm at the American District Telegraph (ADT) Security Systems, Inc. monitored center as described in Section 2.11. The detectors also alarm visibly and audibly at the ship.

2.4.1 Fire Detecting System The originally installed fire detecting system is deactivated, disabled and performs no active function.

Rev. V 20

Updated Final Safety Analysis Report - (STS-004-002) 2.4.2 Fire Detection and Alarm System (FA)

In 2009, a new fire detection and alarm system was installed. The FA system covers occupied areas of the A-deck and above, including associated stairwells, passageways and portions of B-deck aft of the Hold 4 bulkhead. A partial loop covers C deck technical library area and protects these 3 documents storage areas encompassing the breadth of the vessel plus the battery room.

The FA System is a marine listed Fire Alarm and Detection System. The equipment is a multi-loop, Class A computerized and programmable system. The equipment is U.S. Coast Guard-approved, NFPA, and UL listed for marine use.

The FA System is powered by a 120VAC power source via a 120/120VAC isolation breaker. The isolation breaker shall prevent power spikes from damaging the Alarm Panel internal components.

The ship is alarm zoned into specific fire zones (loops). If a device is activated, smoke or pull station, it would require evacuation throughout the ship. The Alarm Panel, Master Pull Station, Battery Back-up Enclosure and other required accessory panels are located in the Security Office on A-Deck.

In accordance with NFPA 72, Fire Safety Codes, a speaker microphone is located at the Alarm panel and speaker strobes are located throughout the ship.

In addition to the four heat sensors installed at the main switchboard, cathodic protection system and each dehumidification system, three heat sensors are installed in Port and Starboard Sideport Entrances - A Deck Frame 150 and Battery Room 4 - C Deck Frame 142.

In these spaces, smoke and/or fumes/vapors could be present under normal conditions and inadvertently activate a Smoke Detector. Smoke Detectors (130) are installed in all other areas of the ship, as required.

Note that in spaces where Heat Sensors are currently installed, if these spaces become sites of industrial activity that generates smoke caused by welding and burning, these areas may also be fitted with a Smoke Detector. Manual Pull Stations (50) are installed in the house at C-deck and above between Frames 102 and 185 at all means of egress.

To ensure the alarm features are available in the event of a loss of power, battery back-ups are installed to maintain the Fire Alarm System in operating condition for a minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in stand-by operation followed by 5 minutes of alarm activation as required by NFPA 72- 1996 or later.

2.5 Lifesaving Equipment All of the life-saving equipment is deactivated, disabled and performs no active function. A limited number of life rings are located on the ship.

2.6 Radio Equipment All of the radio equipment is deactivated, disabled and performs no active function.

2.7 Deck Machinery Most of the principal deck machinery including lifeboat davits and boat winches are deactivated, disabled and performs no active function. The hydraulic anchor windlasses are operable and can be used to lower and raise the anchors which are made up with the anchor chains. The forward mooring capstans are presently disabled, but can be reactivated with minimum effort, if required.

2.8 FloodingAlarms The flooding alarm system is installed and maintained in accordance with standard marine practices for deactivated ships. The ship has flood alarm sensors that will detect flooding in the bilges of engineering Rev. V 21

Updated Final Safety Analysis Report - (STS-004-002) spaces and any spaces that have sea connections. The ship's flood detectors alarm at the ADT monitored center as described in Section 2.11. The detectors also alarm visibly and audibly at the ship.

2.9 DehumidificationSystem The dehumidification (DH) system was installed in 1994 following the ship being placed in protective storage at the James River Reserve Fleet, Ft. Eustis, VA. The DH system lowers the moisture content of the atmosphere of an enclosed area of a ship by the use of machinery. Per the Fleet Reserve Manual, the acceptable range for relative humidity is 35% to 45%. Vent ducting and cabling associated with the DH system generally obstructs the doorways, passageways and ladders throughout ship as is typical in the marine industry for deactivated ships.

The system was deenergized after the crew was established and the ship was routinely manned following the end of the Colonna's Shipyard period in 2007. The crew is expected to note and take action to correct degradation that the DH system would have minimized in an unmanned ship. When the ship is not routinely manned, the DH system is energized.

2.10 Hull Cathodic ProtectionSystem The vessel is equipped with a conventional, marine impressed current cathodic protection system for hull preservation. The system is monitored and maintained as required. It is operated in accordance with standard maritime practices. At times, the system may be supplemented by external cathodic protection as described in the Reserve Fleet Manual.

2.11 Security Systems The intrusion alarm system was improved in 2007 with the addition of new equipment to an existing system under a contract with American District Telegraph (ADT) Security Systems, Inc. ADT will now provide remote monitoring of all installed flooding, fire and intrusion alarms from one of its central ADT monitoring centers. The MOSCAD (Motorola Supervisory Control and Data Acquisition) radio system was deenergized and abandoned in place. The function of the MOSCAD was replaced with an ADT provided cellular phone system that dials out to the continuously manned ADT monitoring center system on receipt of any alarm condition. The ADT center then initiates notification to designated MARAD representatives. The detectors also alarm visibly and audibly at the ship.

2.11.1 Camera Surveillance System A Camera Surveillance System is installed to allow video monitoring of the ship internal and external areas.

2.11.2 Security Entrance Door The normal entrance to the ship is operated by a key fob by activating entry and exit panels. The fob operates to communicate with the door closure via a reader pad using passive radio identification frequency. After opening the door closure ensures the door returns to the closed position. In an emergency, the door can be opened from the inside by an emergency override switch panel. The system records the date and time that the individual holding the fob enters and exits the ship.

2.12 GeneralAlarm The General Alarm System was returned to service after repair and testing. The system is powered by a battery/trickle charger located in the C-Deck Battery Room. The General Alarm is activated by a Manual activation switch in Stateroom A-18.

Rev. V 22

Figure 2-1 Containment, Reactor Compartment and Engine Room (Frames99-148)

Figure 2-2 General Arrangement - Inboard Profile and Decks - Promenade, A and B I I 11, a. . IBOARD PROFILE 1=

PROMENADEDECK -BOAT DECK NVGTO RDEDC TOP2E HOUS

(;a ~L PEC 0" CO a DECK 0-T CTf

Figure 2-3 General Arrangement - Decks - C, D, Flats, Tank Tops and Containment Vessel

-Li1.iU- j CýDECK TM+

'diJ1 1151-

- M--a-4, 16'-S FLATAUX.MACY FLATh I n OFrK O~EP TANNm~

- ----- I -- -~

M.j 24'-6'FLAT SECTONTýOý3NCENTER OF OONTAKNlN V55EL-FR.115 k)

Updated Final Safety Analysis Report - (STS-004-002) 3 SITE AND SITE CRITERIA 3.1 Site Description The licensed site of the NSS is the boundary defined by the ship's hull (see Figures 2-2 and 2-3).

Because the NSS is mobile, the off-ship characteristics of the site vary with the location of the vessel.

During the NSS's operating period, the basic limitations on operations in all ports were described in the Port Operating Criteria which required the development of Port Operating Plans. The specific site characteristics of ports that the NSS entered were evaluated prior to the visit of the ship and documented in a Port Operating Plan. The limitations on operation were based on site analysis assuming the occurrence of the maximum credible accident (MCA) that is associated with reactor startup and operation.

The MCA is a major loss-of-coolant accident (LOCA) from the primary system. The specific LOCA was assumed to be instantaneous, complete, transverse shear of a primary system pipe. The details of the original Port Operating Criteria are described in 3.2 Revision Ill of the FSAR. The details of the LOCA are described in 13.5 of Revision Ill of the FSAR.

Following defueling in Fall 1971 when the reactor was no longer operated, the MCA could have been revised to a worst case defueled accident as there was no longer any possibility of a reactor criticality accident. Regardless, the Port Operating Criteria remained in effect and unchanged as a conservative developmental tool for Port Operating Plans. Technical Specification Change Number 12, dated December 20, 1971, noted that "approval of the proposed changes would permit the ship to be towed, berthed, moored or drydocked in the defueled condition at any domestic location having an approved port entry and docking plan."

Prior to 1976 when the Possession-only License (Amendment 8) was issued, the Port Operating Criteria was an attachment to the Operating License NS-I. Port Operating Plans have been developed and approved by MARAD and forwarded to NRC for information to allow ship movement, berthing, public display and maintenance as needed.

During the ship's operating period, the acceptability of a location was dependent on establishing the low-population zone such that following the MCA, the dose to members of the public would not exceed established limits. Prior to permanent defueling in 1971, the Maximum Credible Accident was a loss of coolant accident assuming continuous operation at 100% for two years. Following defueling in 1971, there has been no possibility of any reactor/criticality accident. Following the 1976 decommission activities to remove the three primary purification system ion-exchangers, their resins and dewatering the primary, auxiliary and secondary systems, no decommissioning activities have been allowed.

Following the final shutdown on November 8, 1970 and the subsequent defueling, the remaining radioactive material is primarily the activated pressure vessel and other primary system components. The activation estimates have dropped dramatically in the subsequent years: 108,496 Ci (1976 calculated using power history and core physics), 13,000 Ci (1994, estimate based on decay of the principal isotopes from the 1976 data) and 452 Ci (2005). Note that the 2005 estimate is lower because it used the actual power history, the results of the intrusive sampling and was calculated using ORIGEN-il computer-based modeling code.

General area dose rates measured during the 2005 radiological and environmental characterization scoping survey are discussed in Chapter 13.

In January 2007, the Safety Evaluation Report for License Amendment 13 noted that:

Given the reduced the inventory of radioactive material onboard, and the low activation estimates and general area dose rates, there is no Credible Accident that could cause the dose to the public to Rev. V 26

Updated Final Safety Analysis Report - (STS-004-002) approach the established dose limit. As a result, the acceptability of the ship's location should be dependent on MARAD safety analysis and prior notification to the NRC. Because MARAD has initiated and directed the towing of hundreds of vessels to layberths, repair yards and the reserve fleets, towing operations incident to relocating the NSS are straightforward and essentially routine.

In approving the Technical Specification change that allowed the NSS to be towed, berthed, moored or drydocked in any U.S. domestic location having a U.S. Maritime Administration approved Port Operating Plan, the NRC concluded:

Given that there is little possibility that any event could lead to a dose approaching the established limit to the public, any difference between the ship being at JRRF or any domestic location with an approved Port Operating Plan is negligible.

Following implementation of License Amendment 13, Port Operating Plans are approved by MARAD and submitted to the NRC for information.

3.2 Ship Movement and Port Operating Criteria 3.2.1 Purpose Notwithstanding the changes described in Section 3.1, the NSS remains a waterborne, mobile nuclear facility. The ability to move the ship and berth it in various locations, for various purposes, is an evolution that remains inherent to the facility. The following Ship Movement and Port Operating Criteria are consolidated from the several licensing precedents cited in Section 3.1, and provide a single source, contemporary guideline for these activities.

3.2.2 Ship Movement Criteria The ship movement criteria define the basic parameters under which routine movement of the NSS may be conducted. Emergency ship movements may be made for a variety of reasons, such as shipboard or shoreside fire, allisions and collisions, security threats, and heavy weather. In such cases the emergency movement of the vessel will be coordinated with, or directed by the United States Coast Guard (USCG),

and will, to the extent practicable, comply with routine ship movement criteria.

1. MARAD will notify NRC in writing in advance of planned movements of the ship. MARAD will promptly notify NRC in the event of an emergency movement of the ship.

2. For any ship movement that involves a change of port, MARAD will review the affected Port Operating Plan(s) in accordance with these criteria, and certify the results to NRC.

3. Ship movements will be governed by applicable USCG regulations and notices. Coordination with local USCG in departure and arrival ports is required for all NSS movements. Similar coordination with and notification to state and local governments and port authorities will be undertaken as required.

4. In general, NSS movements will conform to USCG dead ship tow criteria and conventional marine towing practices and standards. Unless otherwise directed by the USCG, the NSS will transit on normal shipping routes and-channels without limitation.

5. The NSS is a Public Vessel of the United States, and as such may be exempt from certain USCG regulations and requirements. Where such exemptions exist, they will be evaluated jointly by MARAD and the USCG to determine specific applicability to the NSS.

Rev. V .27

Updated Final Safety Analysis Report - (STS-004-002) 3.2.3 Port Operating Criteria These port operating criteria, based on the criteria contained in Revision IV and other licensing precedents, provide basic guidelines for in-port operations and berthing of the NSS in the defueled, non-operating condition. Port Operating Plans, derived from these criteria, shall be prepared and maintained for ports in which the NSS is likely to berthed and/or taken for repairs, maintenance and drydocking.

1. A detailed Port Operating Plan of the port to be visited must be prepared and must be approved by MARAD prior to the ship's visit. The content of Port Operating Plans shall include at least the following:,

A description of the local geography and meteorological / hydrological information; A description of port administrative services and local contact information for applicable service organizations (e.g., United States Coast Guard, Pilots, Tugs, Police and Fire); and,

A description of mooring locations (if known) and transit routes to be used by the ship.

2. Port-specific security and emergency (including radiological) plans must be prepared in advance of the ship's movement. Such plans may be incorporated into the Port Operating Plan, and should demonstrate how licensee personnel and local authorities will be organized and employed to conduct operations that would be necessary after the occurrence of an emergency.

3. Within 30 days of arrival in the port, briefings and walk-throughs of the emergency plans should be conducted with first-responders. When the port visit involves repairs, maintenance or drydocking at a ship repair facility / shipyard; facility-specific plans shall be prepared and implemented upon arrival.

4. Port Operating Plans shall be reviewed at least annually, and updated as required.

5. The criteria for adequate tugs and limitation on visitors are deleted.

3.3 ControlZones Population control zones were no longer required after the Possession-only License was issued.

3.4 Comparison with CentralStation Site criteriain the USA

[Deleted]

Rev. V 28

Updated Final Safety Analysis Report - (STS-004-002) 4 CONTAINMENT 4.1 Introduction The containment comprises a combination of compartments and equipment having the purpose of minimizing release of radioactivity outside the ship following an accident. Prior to and following defuieling in 1971, this system has been the NSS's principal engineered safeguard. During the NSS's operating period, its performance and reliability were vital to safety of the ship and the surrounding environment. The containment is divided into two parts: the primary containment (containment vessel) and the reactor compartment (secondary containment).

4.2 Primary Containment (Containment Vessel) 4.2.1 Function and General Design Criteria The function of the primary containment was to minimize the release of fission products after a serious accident. During the operating phase, the Maximum Credible Accident (MCA) was the accident that would have resulted in the greatest potential hazard to the general public. The entire primary system and all primary components are located inside the pressure containment (Figure 4-1).

A major failure of the primary system (i.e., instantaneous, complete, transverse shear of primary system pipe) in the MCA would have been followed by rapid flashing of the primary coolant water as it was released to the containment. A major portion of the coolant would have escaped and caused the containment pressure to rise. As decay heat was added to the core, lack of adequate cooling would have resulted in an increased core temperature. Ultimately, failure of the cladding would have released fission product to the primary system and containment.

Although the containment must necessarily have access openings and service penetrations where some finite amount of leakage is possible, the pressure containment was designed to hold the largest possible quantity of these fission products. The following criteria were established for the pressure containment:

I1. The containment was sealed at all times during plant operation and no entrance was permitted until the plant had been shut down and the radiation had been reduced to a safe level.

2. Design pressure of the containment vessel was determined by expansion of the primary fluid from the entire primary system and of the secondary fluid from one steam generator. The blow-down occurs through a postulated pipe rupture.

3. No high-temperature piping containing primary water was permitted to penetrate the containment vessel.

4. Electrical containment penetrations were designed to withstand the containment design pressure and temperature.

5. fluid penetrations and piping outside the containment up to and including the first shutoff valve were designed to withstand the containment design pressure and temperature.

6. Any equipment inside the containment vessel that might fail or give trouble during plant operation was installed either in duplicate or in such a manner that the faulty equipment may be isolated from the rest of the plant without hindering the plant operation.

7. The shell and all its penetrations were designed in accordance with American Bureau of Shipping (ABS) Rules, United States Coast Guard (USCG) regulations and American Society of Mechanical Engineers (ASME) requirements.

Rev. V 229

Updated Final Safety Analysis Report - (STS-004-002)

Figure 4-1 Arrangement of Principal Components in Containment I Heat Exchanger (Steam 2 Let Down Cooler 3 Steam Drum (Steam Generator U-tube bundle) Generator) 4 Effluent Condensing Tank 5 Containment Drain Tank 6 Pressurizer 7 Check Valve 8 Primary Pump 9 Gate Valve 4.2.2 Containment Vessel 4.2.2.1 Description The containment vessel (See Figure 4-2) is a horizontal cylinder with hemispherical ends having an outside diameter of 35 feet 4% inches and a length of 50 feet 8/2 inches. It has a centrally mounted vertical cylinder 13 feet 6 inches in diameter extending 16 feet 6 inches above the horizontal cylinder. The gross volume of the containment is approximately 40,000 cubic feet and the net free volume is approximately 31,200 cubic feet. The containment vessel bulkhead thicknesses vary from a minimum of 11/4 inches at the hemispherical ends to 3%

inches in the cylindrical portion of the vessel.

The vessel lies fore and aft on the ship's center line, 18 inches above the inner bottom. It rests in a saddle-shaped support, which is integral with the ship's structure and is secured only at the bottom at the after end of the cylindrical section. This connection provides resistance to Rev. V 30

Updated Final Safety Analysis Report - (STS-004-002) rotational forces and fore-and-aft motion. The support can thus accommodate thermal expansion. Lateral support is provided at mid length by the collision mat, thereby preventing movement due to the roll of the ship. The collision mat itself is composed of alternate layers of one inch steel plates and three inch slabs of redwood. Structural chocks at C-deck level provide lateral support and ensure that the containment remains in place at critical angles, including those in the capsized condition.

The containment has 99 major penetrations and openings through the shell (See Table 4-1, List of Containment Penetrations). Some of these are subdivided into several minor penetrations for electrical or pneumatic service. All of the normal access and maintenance openings are located on the upper half of the vessel. Nearly all of the electrical and fluid penetrations pass through the lower half of the vessel.

4.2.2.2 Structural Design Design of the pressure containment was based on the MCA conditions. Very early in the design phase of the project, the MCA was defined as the instantaneous release of the fluid energy in the primary system while the reactor was at full power. The preliminary calculation for this condition resulted in a peak pressure of 186 psig. The containment shell and all penetrations were designed for this pressure. Later, the accident was specifically defined as a primary system rupture having a flow area equal to the cross-sectional area of the largest primary pipe.

Also included in the energy release was the fluid energy of one main steam generator. Since the rupture area had been defined, the post-accident analysis was made time-dependent and the resultant peak pressure was determined to be 173 psig. The containment vessel hydrostatic test was conducted at this pressure. Later analysis included the effects of a subsequently authorized higher maximum reactor power. It also included the effects of experimental data for post-accident condensing film heat transfer coefficients. The net effect of these two factors on the peak pressure was a reduction to 160 psig. Hence, the shell thickness is approximately 15%

greater than that which is required by Section VIII of the ASME Boiler and Pressure Vessel Code. The design stress of the code is one-fourth of the ultimate tensile stress. The pressure containment detailed design calculations were reviewed and approved by the USCG and ABS.

Maximum stresses in the containment vessel under full shielding and dynamic stress of a 30-degree roll are experienced in the main cylinder rings located 7 feet 6 inches forward and aft of the containment vessel center line and in the inner cylinder rings located 2 feet 6 inches forward and aft of the center line. These stresses were determined by loading each element of the ring sections with vertical and horizontal loads at a 30-degree roll combined with the dynamic effect of this roll. Each ring was analyzed by use of strain-energy equations. The rings were considered stable and statically determinate structures with unit loads applied to each redundant reaction and moment. Unit deflection equations were then obtained, which when equated to zero, resulted in simultaneous equations, whose solutions yielded the reactions and moments.

The final moments in each element of the rings were found and since the section properties of the elements were known, the resulting bending stresses were obtained.

Maximum stresses in the main cylinder ring are 16,000 psi in the inner flange and 8400 psi in the outer shell of the ring section. These stresses occur when support is horizontal (90 degrees to the vertical). Maximum stresses in the cylinder partial ring sections are 20,500 psi in the inner flange of the ring and 6200 psi in the outer shell of the ring section. These stresses also occur when support is horizontal (90 degrees to the vertical).

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The torque developed by a dynamic 30-degree roll, is considered to be carried in the shell of the containment vessel as a shear in the shell. The shearing force resulting from this condition is resisted by the bolts located at the aft external support.

Maximum stresses in the hemisphere rings are not appreciably affected by a 4-degree pitch condition since the longitudinal center of pitch is located near the center line of the containment vessel. The stress in these rings is 20,000 psi for a static condition at a point 90 degrees from the vertical.

The external structural supports at 90 degrees to the vertical port and starboard are stressed to 6000 psi by the reactions from the cylinder rings. These stresses occur in the D-deck transverse members during a 30-degree roll condition.

The external foundations of the containment vessel were designed to resist the reactions from the hydrostatic test. These reactions exceeded those experienced during a 30-degree roll. Thus, the containment vessel foundations are suitable for all loading conditions, since maximum reactions due to the hydrostatic test are the most critical reactions experienced by the external foundations.

All internal equipment foundations are designed to stresses less than 12,500 psi for the combined roll-and-pitch condition.

4.2.2.3 Fabrication and Strength Testing The cylindrical section, the two hemispherical heads and the cupola were fabricated as four separate units. All welds in each section were preheated and the roots and fillets were inspected by Magnaflux methods. After completion of welding, all welds were radiographed and the four sections were stress-relieved. The sections were then lowered into the reactor compartment and welded together manually by the same procedures used for the individual pieces. Since the final assembly welding of the sections was on sections of plate whose thickness was less than 11/22 inches, no final stress relieving was required. Before installation of components in the containment, the completed vessel was hydrostatically tested at 173 psig with the penetrations capped or plugged closed.

The fabrication and welding procedures were reviewed by the USCG and the ABS. Site inspectors were present during fabrication and testing of the vessel to ensure compliance with the approved procedures.

4.2.3 Containment Vessel Penetrations The containment penetrations, the containment isolation valves and their location and method of operation are listed in Table 4-1. No penetration has been tested since final shutdown November 8, 1970.

There are six openings which can be used for access or equipment removal. All but one (the double-door air lock) of these openings are bolted closures. The 42 inch double-door air lock permitted entering the containment without breaking containment integrity. In addition to the air lock, a 42 inch diameter bolted hatch is located forward of the cupola. This hatch can also be used for access and servicing of small equipment when the plant is shutdown. A CV entry shield plug can be placed on top of the upper airlock door and on top of the bolted hatch.

A 13 1/22 feet, full-diameter hatch at the top of the cupola is provided for major maintenance and refueling operations. A 51/2 /2feet diameter hatch is located in the top of the cupola for control rod Rev. V 32

Updated Final Safety Analysis Report - (STS-004-002) drive removal. Two combination, 18" x 24" manways and flooding hatches are located on the starboard side in the lower quarter of each hemispherical head.

There are 99 service penetrations that carried air, water, steam, hydraulic fluid and electrical power through the containment vessel shell. The inerting piping in the CV cupola head was disconnected outside and valves closed so that the "integrity" of the vessel was maintained until the characterization activities in 2005 required the CV be opened and ventilated using the nitrogen purge piping for supply and exhaust. Supply air was provided from the ship's interior dehumidified air distribution ducting when. The discharge of the ventilation cycle is through a HEPA filter located in the upper level reactor compartment. The modified system was left in place after the characterization activities were complete.

Piping penetrations less than 2 inches in diameter utilize a simple nozzle, while larger pipes use a nozzle with a reinforcing doubler. The penetrations for the 8 inch diameter main steam line from the top drum of each main steam generator are sealed off with a special 14 inch diameter expansion joint, which is welded to the pipe at the top and to the containment at the bottom.

There are only two other service penetrations which are larger than the main steam lines: the 10 inch air-purging penetrations. However no special provision for attachment is made except for the reinforcing doubler, since these pipes are essentially always at the same temperature as the containment.

Electrical penetrations are grouped together in boxes at the point where the cables penetrate the containment. The number of subpenetrations in each major penetration varies from 8 to 25 depending on service, conductor size and location of equipment being served. Leakage occurred through the original penetrations. The original electrical penetrations were replaced with fused glass-to-steel fittings. The electrical penetrations were individually tested at 286 psig with helium.

4.2.4 Containment Auxiliary Features In addition to the pressure vessel itself, many auxiliary features were required to fulfill both normal and emergency functions. During normal operation the atmospheric conditions both inside and outside the pressure containment was maintained in a suitable range for the installed equipment.

This was accomplished inside by the Containment Cooling (CC) system, which is an arrangement of air distribution ducts, duplicate two-speed circulating fans and cooling coils. The circulating fans were operated and supplied with power from either the main control room or the emergency cooling panel on the navigation bridge deck. Associated with the CC system was the necessary instrumentation to measure pressure, temperature and humidity. By maintaining a slight positive pressure in the containment, any large change in containment integrity was brought quickly to the operator's attention. This system maintained an average internal containment air temperature of about 1300F. As a backup to this system, when the normal cooling water supply was unavailable, a third cooling coil was available as a part of the Emergency Cooling (DK) system. This coil was supplied directly with sea water.

The atmospheric conditions in the reactor compartment were maintained by the Reactor Space Ventilation (RSV) system. Air that had leaked into the reactor compartment was discharged through filters and up the 90 foot mast forward of hold 4.

All of the CC, DK and RSV equipment is deactivated, disabled and performs no function. Any ductwork or other components removed, as part of removing the RCP motors in 1971 or removing the primary purification ion exchangers (demineralizers) in 1976, was not replaced.

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Following a major pipe rupture inside the containment, it was desirable to reduce the post-accident pressure. The following design features permitted this pressure reduction:

Either the normal or emergency cooling coils in the CC system were used. Heat was removed by forced air circulation if the fans were operable or by natural air circulation if they were not.

The possibility of forced circulation was enhanced because the fans at half speed were capable of operating in a steam air atmosphere below 100 psi. Either forced or natural circulation could have been increased after the accident by the automatic opening of two large side plates in the cooling ductwork located just below the fans.

" Pressure reduction was accomplished by spraying water from the fire main into the containment through the CO 2 injection system.

Heat removal and pressure reduction was accomplished by partially flooding the lower reactor compartment so that heat flows through the containment shell into the flooding water.

Following the MCA, the possibility existed that radioactivity could escape through some of the fluid penetrations. In the case of most penetrations, this escape would have occurred only if the MCA broke a pneumatic, hydraulic, or water pipe inside the containment to permit radioactive fluid to leave the containment through the pipe. However, this escape was prevented because each fluid line that is in service during normal operation contained either a check valve or a quick closing shutoff valve. With only two exceptions, these quick closing valves are tripped closed if containment pressure reaches 5 psig. The two exceptions were the two heavy-walled, 8-inch main steam lines, which were secured from the main control room because of the possible desire to continue operating the ship's service turbine generators for a brief period after the accident.

Because the containment was inaccessible during operation, hydrogen content of the containment atmosphere, neutron level outside the primary (neutron) shield tank and the gamma level near the containment air lock plus temperature, pressure and humidity parameters could be monitored.

The use of hydraulic oil in the containment meant that there was a possibility of fire if a combustion-supporting atmosphere is present. When the NSS began operations some leaking of hydraulic fluid from the control rod drive system into the containment was experienced. Since the fluid was flammable and under some conditions was subject to flashing when spilled on the hot vessel surfaces, the containment atmosphere was inerted with a nitrogen (ND) blanket to eliminate the possibility of combustion (less than 10% oxygen). There were other occasions of leaking hydraulic fluid after the containment was inerted, but no combustion problems resulted. A complete spare charge of nitrogen was kept aboard the ship to recharge the containment if an outside supply was unavailable. The containment inerting arrangement is shown in Figure 4-2.

Another feature of the containment design provides assurance that the containment would not collapse should the ship be involved in an accident which results in sinking. On the lower half of the vessel are two spring-return, bolted, flooding hatches which open and permit sea water to enter the containment if the ship sinks to a depth of 100 feet or more. When the containment is flooded and the pressures are equalized, the hatches re-close, thereby preventing escape of water from the containment through the flooding hatches. These hatches remain installed but have not been verified functional since final shutdown in November 1970.

For a sinking accident in shallow water, the containment has four 2" salvage connections on the cupola head. Shutoff valves are located immediately adjacent to the vessel on each of these connections and blank flanges are installed on the open end of each valve.

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Figure 4-2 Containment Inerting Schematic CM-6 1V P

jShoreside 0 00 N2 Trailer Ship's Service Air CM-50V CM-60V N Compressors CM-64V L N2 Bottles Engine Room Hold #4 Lower Level Fan RS-P1 Exhaust to Atmosphere Fan RS-P2 WL-12V WL-13V WL-1 12V Reactor Compartment 4.2.5 Containment Performance Requirements No leak rate tests on the pressure containment have been performed since final shutdown November 8, 1970.

4.3 Reactor Compartment (Secondary Containment) 4.3.1 General Description As shown in Figures 4-3 and 4-4, the reactor compartment surrounds the containment vessel and forms a secondary containment around the pressure containment. The concrete, lead and polyethylene form the secondary shield. The volume enclosed by the concrete is the lower reactor compartment or secondary shield area. This shield is described in 5.3.2.

The principal auxiliary system associated with this compartment is the reactor space ventilation (RSV) system. It was installed during the spring 1963 outage. The system consists of two physically independent sections. One section was for heat removal from the upper part of the reactor compartment; the other section was for exhausting air from the reactor compartment.

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Figure 4-3 Reactor Compartment REACTOR COMPARTMENT , , UPPER REACTOR BOUNDARFY NEGATIVE PRESSURE ,AREA

* '*~l COMPARTMENT

,"TO' AIR INLET I-- I~~LEAODI AIR INLETTO I* '° ° *"s TOP OF MAST VALVE -EXHAUST FANS AND CO NC R E T E SE CONDAR Y DUCTWORK SIECONDARY 4.3.2 Cooling Section The cooling (or upper reactor compartment recirculation) part of the RSV system is deactivated, disabled and performs no active function. It is located entirely in the reactor compartment. Two air intakes were provided. The recirculating air intake is on the port side at the A-deck level. The fresh air intake was supplied by fan 1-105-1 via regulating valve RS-40V. The recirculation fans (one is a spare) took suction on the intake manifold and discharge past the salt-water-cooled air cooling coils into a common header. This header distributed the air throughout the upper part of the reactor compartment.

Any ductwork or other components removed, as part of removing the RCP motors in 1971 or removing the primary purification ion exchangers (demineralizers) in 1976, was not replaced.

The water supply for the cooling coils (RS-CI and RS-C2) was from the sea water pumps of the intermediate cooling water (CW) system. A booster pump RS-P5 provided additional head for the cooling water when required.

4.3.3 Exhaust Section 4.3.3.1 Function The exhaust system of the RSV system is deactivated, disabled and performs no active function.

It was part of the engineered safeguards for the ship. Its primary purpose was to retain any particulate and halogen fission products which may leak from the containment vessel. The effectiveness of its filter system was tested periodically to demonstrate desired efficiency for iodine and particulate removal.

The exhaust section of the RSV system maintained the reactor compartment at a pressure of approximately 1 inch wg below the B-deck corridor pressure. The reactor compartment pressure was below that of the surrounding shipboard areas and the containment vessel. This arrangement ensured that any containment vessel leakage was exhausted via the filter-adsorber banks in the RSV system and not to the surrounding shipboard areas. The reactor compartment Rev. V 36

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4.3.3.2 Containment Vessel Purging The containment vessel was purged via the exhaust section of the RSV system. Purging was done by opening the containment vessel purge valves (WL-12V and WL-13V) and closing a damper in the suction of the exhaust section of the RSV system. Air was thus drawn from the reactor compartment into the containment vessel and then to the exhaust section of the RSV system.

4.3.3.3 Suction Ductwork and Lower Plenum The exhaust section of the RSV system took suction in the lower part of the reactor compartment via a single duct to the lower plenum. Separate connections on the lower plenum lead to the two filter-adsorber banks. Motor-operated valves (RS-3V and RS-4V) adjacent to the lower plenum provided flow control. Controls for these valves are located nearby. Normal operation was with one of the filter-adsorber banks in service and the other on standby.

4.3.3.4 Filter Units The filter units consist of demisters, roughing filters, absolute filters and iodine adsorber assemblies. The filters and adsorbers were each designed for a specific purpose. Each filter and adsorber is described below.

The demister was the first process unit downstream of the inlet valves and is provided for the removal of any entrained water in the exhaust gases. The demister consists of a mat of Teflon yarn on a stainless steel core. This type of construction had been extensively tested at the Savannah River Laboratories (SRL).

The roughing filter follows the demister. Its purpose was to remove particulate matter from the exhaust stream. The filtering medium consists of a glass wool mat arranged in a pleated pattern for greater surface area.

The absolute filter is located downstream of the roughing filter. Its purpose was to remove very fine particulate matter which passes through the demister and roughing filter. The filtering medium is a glass sheath that is arranged in the form of deep pleats for additional surface area.

The absolute filters were tested quarterly and prior to each port entry for particulate matter removal efficiency. This test was performed using an aerosol produced by bubbling air through the liquid dioctyl phthalate (DOP). The aerosol was carried along in the RSV system by the exhaust air stream and the DOP concentration was determined on both sides of the filter-adsorber bank. Each absolute filter was tested many times and the efficiency was consistently indicated to be more than 100 times higher than the required 99.9%.

The iodine adsorber assembly follows the absolute filter. This unit consists of three parts mounted in a common housing. The first and last parts of the assembly are identical and consist of mats of silver-plated copper ribbon. This material, made of 4% silver and 96% copper, is compressed to a density of 27 lb/ft3 and is 4 inches thick. The function of this adsorber is to remove halogens from the exhaust stream by chemical combination with the silver.

The middle part is a charcoal adsorber. Granular activated charcoal is contained in 10 horizontal beds, each of which is 2 feet wide, 4 feet long and 1 inch thick. The charcoal is held in place by perforated plates. Baffles at the ends of the trays ensure parallel flow through all 10 beds.

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