ML20217C200
| ML20217C200 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 06/28/1991 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20217C197 | List: |
| References | |
| NUDOCS 9107120104 | |
| Download: ML20217C200 (316) | |
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NITACllMENT TO P 91217 REVISION TO Tile FORT ST. VitAIN PROPOSED DECOMMISSIONING PIAN DATED JUNE 28, 1991 l t t l l l p 3 f -( *( )( ([3 y,7 PDR
_ _ _ .- -- .__ - =- _ _ . -. _- -. e Attachment to P-91217 , July 1, 1991 DIRECTIONS FOR ENTERING CHANGES TO THE FORT ST. VRAIN PROPOSED DECOMMIS$10NING PLAN l As discussed in the cover letter, this attachment provides a l revision to the Proposed Decommissioning Plan and is based on PSC's commitment to update the PDP to incorporate PSC res>onses to the NRC Request for Additional Information, dated February 8, 1991 (G 91020), This revision also incorporates additional details provided in the Fort St. Vrain Detailed Decommissionina Cost i Estimate, dated June 6, 1991 (P-91198) and the PCRV Activation 1 Analysie Verification, dated April 15,1991(P91137). Changes from the original Proposed Decommissioning Plan (submitted on November 5, 1990, in PSC letter P 90318) have been identified in l the following pages by underlining to facilitate ease of review. The only exceptions to this approach are Sections 3.2 and 3.3, which have been replaced in their entirety and are not underlined. The reviewer is requested to- either (1) replace the pages in their l copy of the Fort St. Vrain Proposed Decommissioning Plan with the i attached revision (preferred), or (2) file this letter (with l attachment) in total with the original copy of the Proposed Decommissioning Plan. Final replacement should be consistent with the List of Effective Pages provided immediately following this page. The List of Effective Pages should be alaced in the Proposed Decommissioning Plan immediately following tae Cover Graphic page and preceding the Table of Contents, i w-- m - ~ - -
6/28/91 PROPOSED DECOMMISSIONING PLAN LIST Of effec 11VE PAGES LIST OF EFFECTIVE PAGES SECilDB EFFECTIVE PAEji DATE OF CHANGE Cover Page Originnt (11/5/90) List of (f f ective A ard g 6/28/91 Pages table of Contente i through Riv 6/28/91 1.1 1.1 1 through 1.1 4 6/28/91 1.2 1.2 1 through 1.2 3 6/28/91 1.3 1.3 1 through 1.3 1 6/28/91 1.4 1.4 1 through 1,4 3 6/28/91 1.5 1.5 1 through 1.5 4 6/28/91 1.6 1.6 1 through 1.6 1 6/28/91 2.1 2.1 1 through 2.1 2 Original (11/5/90) 2.2 2.2 1 through 2.2-14 6/26/91 Tables 2.2 1 through 2.2-3 Original (11/5/90) Figures 2.21 through 2.2 31 Originot (11/5/90) 2.3 2.3 1 the mgh 2.3 38 6/28/91 tables 2.3 1 and 2.3 2 6/28/91 Figures 2.3 1 through 2.3 4 Original (11/5/90) Figures 2.3 5 through 2.3 10 6/28/91 figures 2.3 11 through 2.5 14 Original (11/5/90) figure 2.3 15 6/28/91 2.4 2.4 1 through 2.4 6 6/28/91 Figure 2.4 1- 6/28/91 2.5 2.5 1 through 2.5 10 6/28/91 Table 2.5 1 6/28/91 Tables 2.5 2 ard 2.5 3 original (11/5/90) figure 2.5 1 Original (11/5/90) figure 2.5 2 6/28/91 2.6 2.6 1 through 2.6 4 6/28/91 2.7 2.71 ard 2.7 2 6/28/91 3.1 3.1 1 thtough 3.1 18 6/28/91 fable 3.1-1 Original (11/5/90) Table 3.1 2 6/28/91 Tables 3.1 3 and 3,1 4 Original (11/5/90)
- Tables 3.15 ard 3.16 6/28/91 figures 3.1 1 thewgh 3.130 Original (11/5/90) 3.2 3.2*1 through 3.2 76 6/28/91 Table 3.2-1 6/28/91 figures 3.2+1 ord 3.2 2 6/28/91 3.3 3.3 1 through 3.3 16 6/28/V1 Tables 3.31 ord 3.3 2 Original (11/5/90)
Tables 3.3 3 ard 3.3 4 6/28/91 Tables 3.3 5 ard 3.3-6 Original (11/5/90) figure 3.3 1 */28/91 A
PROPOSED DELOMMISS10NING PLAN 6/28/91 TABLE Of LON1EN15 3.4 3.4 1 thr wah 3.4 22 6/28/91 , inbles 3.41 ard 3.4 2 6/28/91 l Tables 3.4 3 throwh 3.4 5 Ortstrat (11/5/90) I figures 3.41 ard 3.4 2 6/28/91 figures 3.4*3 ard 3.4 4 Original (11/5/90) 3.5 3.5 1 thr wgh 3.5 6 6/?t/91 l l 4 4 1 through 4 10 6/28/91 i 5 51 thrwgh 5 8 6/28/91 6 6 1 through 6 3 Original (11/5/90) 7 7 1 through 7 4 6/28/91 7 $ thrcugh 7 14 OrIgital (11/5/90) 8 81 through 8 5 original (11/5/90) 9 . 9 1 through 9+6 6/28/91 10 10 1 through 10 2 6/28/91 1 A;4erdia 1 1 1 throup l 13 Original (11/5/90) Agerdis 11 1 thr.,up 40 (Et DEC 0010, Rev 8.) Ortstral (11/5/90) 4pendta B (26 pages) Original (11/5/90) ' Asterdia C (58 pages) Original (11/5/90) Attachment 1 (9 pages) Original (11/5/90) Attachment 2 (i page) Original (11/5/90) , i i ( B 1 1
6/28/91 PROPOSED DECOMMISS10!41NG PLAf4 1ABLE Of CONTENTS PROPOSED DECOMM1$$10NING PL AN FOR FORT ST. VRAIN NdCLEAR GENERATING STATION TABLE OF CONTENTS
- 1.
SUMMARY
Of PLAN 1.1 Description of Decommissioning Plan and Decommissioning Alternative....................... 1.1 1 1.2 Major Tasks, Schedules a id Activities. . . . . . . . . . . . . 1.2 1 1.3 Fixed Price and Availability of funds............. 1.3 1 1.4 Regulatory Basis for Administration of the Decommissioning Plan.............................. 1.4-1 1.5 Decommissioning Controls During the Transition Period Prior to Approval of the Proposed Decommissioning Plan.............................. 1.51 1.6 References for Section 1.......................... 1.6-1
- 2. Cil01CE Of DECOMMISSIONING ALTERNAllVE AND DESCRIP110N Of ACTIVITIES 2.1 Decommissioning Alternative....................... 2.1-1 2.2 facility Description 2.2.1 General Description...................... 2.2 1 2.2.2 Prestressed Concrete Reactor Vessel (PCRV)and Internal Component s . . . . . . . . . . . . 2.2 2 2.2.3 Balance of Plant Contaminated Components............................... 2.2-4 2.2.4 Site Characteristics..................... 2.2-8
?.3 Decommissioning Activities, Planning and Exposure Estimates 2.3.1 Introduction............................. 2,3-1 2.3.2 Technical Approach Selection............. 2.3 2 2.3.3 PCRV Dismantlement and Decontamination... 2.3 5 2.3.4 Contaminated Balance of Plant System Dismantlement and Decontamination........ 2.3-24 2.3.5 Decommi ssioning Schedule. . . . . . . . . . . . . . . . . 2.3-34 2.3.6 Occupational Exposure Estimate. . . . . . . . . . . 2.3-35 i
i PROPOSED DECOMMISSIONING PLAN 6/28/91 TABLE Of CONI [NTS 2.4 Deconnissioning Organization and Responsibilities 2.4.1 PSC Commitment........................... 2,4 1 2.4.2 PSC Occonnissioning Organization and functions............................ 2.4 1 : 2.4.3 PSC Corporate Vice President. Nuclear 0porations....................... 2.4 2 2.4.4 Project Manager for Occonnissioning...... 2.4 2 2.4.5 Project Assurance Manager................ 2.4 2 2.4.6 facility Support Mana 2.4 3 2.4.7 Operations Manager...ger................. .................... 2.4-4 2.4.8 Cngineering Manager (Deconnissioning).... 2.4 4 2.4.9 Deconnissioning Safety Review C o nn i t t e e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4 4 2.5 Contractor Responsibilities 2.5.1 Westinghouse Team Organization and functions................................ 2.5 1 2.5.2 Westinghouse Team Scope nf Work.......... 2.5 1 i 2.5.3 Organization of the Westinghouse Team.... 2.5 1 2.5.4 Westinghouse Team Qualifications and Experience........................... 2.5 7 2.6 Training Program 2.6.1 General Eniployee Training. . . . . . . . . . . . . . . . 2.6-1 2.6.2 Radiation Worker Training................ 2.6-1 2.6.3 Specific Job Training.................... 2.6 2 2.6.4 Non Radiation Worker Indoctrination...... 2.6-3 2.6.5 Radiation Protection Staff Training...... 2.6 3 2.6. 6 . Training Records,........................ 2.6 3 2.7 References for Section 2.......................... 2.7 1
- 3. PROTECTION Of OCCUPATIONAL AND PUBLIC llEALTH AND SAFETY 3.1 facility Radiological Status 3,1,1 facility Operating llistory............... 3.1 1 3.1.2 Radiation Sources........................ 3.1-2 3.1.3 Current Environmental Radiological Status................................... 3.1 4 3.1.4 Radionuclide inventory................... 3.1 7 3.1.5 Initial Site Characterization Plans...... 3.1-16 11 l
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6/28/91 PROPOSED DEC0KMISS10NING PLAN TABLE Of CONTENTS 3.2 Radiation Protection Program 3.2.1 Introduction............................. 3.2-1 3.2.2 Management Policy........................ 3.2 1 3.2.3 Radiation Protection Organization and functions............................ 3.2 4 3.2.4 Radiation Protection Training and Qualification............................ 3.2 17 3.2.5 Dose Control............................. 3.2 27 3.2.6 Radioactive Materials Control............ 3.2-47 3.2.7 Surveillance............................. 3.2 56 3.2.8 Instrumentation.......................... 3.2 63 3.2.9 Review and Audit......................... 3.2-67 3.2.10 Radiation Protection Performance Analysis................................. 3.2-68 3.2.11 Radiation Work Practices................. 3.2-71 3.3 Radioactive Waste Management 3.3.1 Spent fuel Disposal...................... 3.3-1 3.3.2 Radioactive Waste Processing............. 3.3 2 3.3.3 Radioactive Waste Disposal............... 3.3-10 3.3.4 Disposal of Non-Radioactive Waste. . . . . . . . 3.3-15 3.4 Accident Analysis 3.4.1 Introduction and Description of Decommissioning Accidents................ 3.4 1 3.4.2 Assumptions.............................. 3.4 4 3.4.3 Dropping of Contaminated Concrete Rubble Accident.......................... 3.4 5 3.4.4 Conversion Construction Accident Near PCRV Dismant1ement.................. 3.4 7 3.4.5 Heavy Load Drop Accident................. 3.4 8 3.4.6 fire..................................... 3.4-11 3.4.7 Loss of PCRV Shielding Water Accider.t... . 3.4-13 3.4.8 Loss of Power............................ 3.4-15 3.4.9 Natural Disasters........................ 3.4-17 3.4.10 Summary.................................. 3.4-21 3.5 References for Section 3.......................... 3.5 1
- 4. FINAL RADIATION SVRVEY PLAN 4.1 Introduction...................................... 4-1 4.2 Fi n al Rel e a se Cri t eri a . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.3 S u rvey Me t hodol ogy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.4 Instrumentation................................... 4-6 4.5 Dccumentation..................................... 4-7 4.6 Quality Assurance................................. 4-8 4.7 Independent Verification.......................... 4-8 4.8 References for Section 4.......................... 4-10 tii
i 1 PROPOSED DECOMMISSIONING PLAN 6/28/91 , TABLE OF CONTENTS
- 5. DECOMMISSIONING flXED PRICE CONTRACT AND FUNDING PLAN ;
5.1 Decommissioning Contract.......................... 51 5.2 Major Assumptions, Bases, and Scope of fixed Price Contract........................... 52 5.3 Decommissioning Cost Dreakdown.................... 54 5.4 Decommissioning Funding Plan...................... 55 5.5 Updates to the Decommissioning Funding Plan....... 5-5 5.6 References for Section 5.......................... 57
- 6. DECOMMISSIONING TECHNICAL AND ENVIRONMENTAL SPECIFICA110NS 6.1 Introduction...................................... 61 6.2 DTS Limits and Controls........................... 61
- 7. DECOMMISSIONING QUALITY ASSURANCE PLAN 7.1 Policy Statement.................................. 7-1 7.2 Introduction...................................... 7-1 7.3 Organization...................................... 72 7.4 Quality Assurance Plan............................ 73 7.5 Design Control.................................... 75 7.6 Procurement Document Control...................... 7-5 7.7 Procedures and Drawings........................... 76 7.8 Document Control.......................... ....... 77 7.9 Control of Purchased Material. Equipment and Services...................................... 78 7.10 Identification and Control of Materials, Parts and Components.............................. 7-9 7.11 Control of Special Processes...................... 7-9 7.12 Inspection........................................ 7 10 7.13 Test Control...................................... 7-10 7.14 Control of Measuring and Test Equipment........... 7-11 7.15 Handling, Storage, Shipping....................... 7-11 7.16 Inspection, Test, and Operating Status............ 7-12 7.17 Nonconforming Materials, Parts or Components... .. . 7-12 7.18 Corrective Action................................. 7-12 7.19 Quality Assurance Records......................... 7 13 7.20 Audits............................................ 7-13 l
l 8. DECOMMISSIONING ACCESS CONTROL PLAN 8.1 8 asis for Access Control Plan..................... 8-1 8.2 Site Access Control Organization.................. 82-8.3 Access Control Physical Security Measures......... 8-2 8.4 Communications.................................... 8-3 8.5 Procedures........................................ 83 l i iv l 1
6/28/91 PROPOSED DECOMMISSIONING PLAN TABLE Of CONTENTS ,
- 9. DECOMMISSIONING EMERGENCY RESPONSE PLAN 9.1 Introduction and Regulatory Basis................. 91 9.2 Dermmmissioning Emergency Response Plan Scope..... 91
- 10. DECOMMISSIONING f!RE PROTECTION PLAN 10.1 Introduction...................................... 10 1 APPENDICES
- 1. WESilNGH0VSE TEAM SCOPE Of WORK
- 11. FORT ST. VRAIN ACTIVATION ANALYSIS y
PROPOSED DECOMMISSIONING PLAN 6/28/91 TABLE Of CONTENTS LIST Of TABLES 2.2 1 Public Water Supplies Within Thirty Miles of fort St. Vrain 2.2-2 Hours at Each Wind Speed and Direction 2.2-3 frequency of Distribution for Each Wind Speed and Direction 2.3 1 Estimated Contact Dose Rates for Graphite Blocks 2.3-2 Projected _ Person Rem Exposure for the fort St. Vrain Decommissioning Project 2.5 1 Relevant Westinghouse Team Experience 2.5 2 facilities Approved for Unrestricted Release 2.5 3 Waste Handling and Packaging Experience 3.1 1 Radiological Survey Summary 3.1-2 Activation Analysis Results 3.1-3 PCRV Dose Rates in Air at 5 Years After Shutdown 3.1-4 Estimated Plateout Concentration on Major Primary Circuit Components at E005 3.1 5 Integrated Plateout in Each Primary Circuit Component at E005 3.1-6 Estimated Curio Total At fort St. Vrain 3__. 2 1 , Typical fort St. Vrain Decommissioni;ig Monitoring hLslruments 3.3-1 KRV Waste Classification and Volume Reduction 3.3-2 Contaminated Balance of Plant Waste Classification and Volume Reduction 3.3-3 PCRV Waste Volume Estimates 3.3-4 B0P Waste Volumes Estimates 3.3 5 Waste Class and Container Type for PCRV 3.3-6 Waste Class and Container Type for B0P 3.4-1 Summary of Accident Scenarios 3.4 2 Doses to an Individual Off-site from Postulated Accidents 3.4-3 Curie Totals in Activated PCRV Concrete 3.4-4 Percentage Contribution of Activation Products in first 6 Inches of Top Head Concreto 3.4-5 Waste Volume / Activities Estimates for the PCRV l vi
6/28/91 PROPOSED DECOMMISSIONING PLAN TABLE Of CONTENTS LIST OF FIGURES 2.2-1 for St. Vrain Plot Plan 2.2 2 Reactor and Turbine Building - Plan View 2.2 3 Reactor and Turbine Building - Elevation View 2.2 4 fuel Storage Building : 2.2 5 Prestressed Concrete Reactor Vessel (PCRV) 2.2 6 PCRV General Configuration 2.2 7 Core Arrangement Elevation Section 2.2-8 Thermal Barrier Arrangement 2.2-9 Core Support Arrangement 2.2 10 fuel Handling Machine 2.2-11 Reactor O 31ation Valve 2.2-12 Auxiliary Transfer Cask and Associated Auxiliary Equipment 2.2 13 fuel Storage Wells 2.2 14 Reactor Plant Arrangement - Equipment Storage Wells 2.2-15 Hot Service Facility 2.2 16 Reactor Plant Arrangement - Refueling floor layout 2.2-17 Helium Circulator Auxiliary Equipment (System 21) 2.2-18 Helium Purification System Auxiliaries 2.2-19 HeliumStorageSystem(System 24) 2.2 20 Reactor Plant Cooling Water System (System 46)
-2.2-21 Purification Cooling Water System (System 47) 2.2 22 Decontamination System (System 61) 2.2 23 Radioactive Liquid Waste System (System 62) 2.2-24 Radioactive Gas Waste System (System 63) 2.2-25 Reactor Building Drain System (System 72) 2.2 26 Reactor Building Ventilation System (System 73) 2.2-27 Area Within Thirty Miles of Fort St. Vrain 2.2-28 Subsurface Geology Surrounding The Site 2.2 29 Estimated Bedrock Contours 2.2-30 Estimated Water Table Contours 2.2-31 Major Tributaries and Irrigation Ditches l 2.3 1 Helium Circulator Installation i 2.3-2 Steam Generator Module l
2.3 3 PCRV Lower Plenum 2.3 4 PCRV Water Cleaning / Clarification System 1 2.3 5 PCRV Water Clarification System 2.3 6 PCRV Top Head Cutting Arrangement 2.3-7 PCRV Cutting Configuration - Inserting the Diamond Wire 2.3 8 PCRV Cutting Configuration - Inclined Core Drilled Holes vli l
PROPOSED DECOMMISS10NING PLAN 6/28/91 TABLE OF CONTENTS 2.3 9 PCRV Cutting Configuration - Vertical Sectioning Cuts 2.3 10 Removal of Remaining PCRV Top Head Concrete 2.3-11 Elevation View of PCRV Work Area 2.3-12 Steam Generator Shipping Container 2.3-13 PCRV Beltline Concrete - Horizontal Cuts 2.3-14 PCRV Beltline Concrete - Vertical Cuts 2.3-15 Schedule of Decommissioning Tasks 2.4-1 PSC Decommissioning Organization 2.5-1 Westinghouse Team Organization Chart 3.1-1 Compactor Building Radiation Survey 3.1-2 Radiochemistry Laboratory Radiation Survey 3.1-3 Turbine Building Radiation Survey - Level 5 (Elev. 4791') 3.1-4 Turbine Building Radiation Survey - Level 6 (Elev. 4811') 3.1 5 Turbine Building Radiation Survey - Level 7 (Elev. 4829') 3.1-6 Turbine Building Radiation Survey - Levels 8, 10 & 11 (Elev. 4846', 4864', 4884') 3.1-7 Turbine Building Radiation Survey - Level 12 & 13 (Elev. 4904', 4921') 3.1 8 Reactor Building Radiation Survey - Level 1 (Elev. 4740') 3.1-9 Reactor Building Radiation Survey - Level 2 (Elev 4756') 3.1-10 Reactor Building Radiation Survey - Level 3 (Elev. 4771') 3.1-11 Reactor Building Radiation Survey - Level 4 (Elev. 4781') 3.1-12 Reactor Building Radiation Survey - Level 5 (Elev. 4191') 3.1-13 Reactor Building Radiation Survey - Level 6 (Elev. 4816') 3.1-14 Reactor Building Radiation Survey - Level 7 (Elev. 4829') 3.1-15 Reactor Building Radiation Survey - Level 8 (Elev. 4859') 3.1-16 Reactor Building Radiation Survey - Level 9 (Elev. 4859') 3.1-17 Reactor Building Radiation Survey - Level 10 (Elev. 4864') 3.1-18 Reactor Building Radiation Survey - Level 11 (Elev. 4881') Refueling Floor 3.1-19 Reactor Building Radiation Survey - Levels 10 & 11 (Elev. 4864' & 4881') 3.1-20 Turbine and Reactor Building Elevations 3.1-21 Location of Site Trailers 3.1-22 PCRV and Internal Components 3.1-23 Core Arrangement [ 3.1-24 -Class-A Insulation and PCRV Liner 3.1-25 Central Column Metal Clad Reflector 3.1 Side Column Metal Clad Reflector 3.1-27 Region Constraint Devices (RCDs) 3.1-28 Lower Orifice-Valve Assembly 3.1-29 Core Support Arrangement 3.1-30 Class C Insulation viii l
6/28/91 PROPOSED DECOMMISSIONING PLAN TABLE Of CONTEN1S 1 3.2-1 Westinghouse Team Radiation Protection Organization Chart 3.2 2 Radiation Protection Program Manual Structure 3.3 1 Estimated Tritium inventory in PCRV Water System 3.4-1 Whole Body Exposure at EAB 3.4 2 PCRV Work Area - Elevation View 3.4-3 Large Scale Regionalization for Tornado Risk Analysis 1 l ix
l l PROPOSED DECOMMISSIONING PLAN 6/28/91 TABLE Of CONTENTS COMMONLY USED %CRONYMS AEC Atomic Energy Commission ALARA As low As Reasonably Achievable ANSI American National titandardt Institute A00 Anticipated Operational Occurrences ASTM Amet ican Society of Testing and Materials ATC Auxiliary Transfer Cask BNL Battelle Northwest Laboratories BOC Beginning of (fuel) Cycle BOP Balance of Plant COOH Colorado Department of Health CEBAf Continuous Electron Beam Accelerator facility CFR Code of federal Regulations Ci Curie CPI Consumer Price Index CPM Counts per minute CPVC Colorado Public Utilities Commission CRD Control Rod Drive CRDOA Control Rod Drive and Orifice Assembly CSF Core Support floor D/D Decontamination and 0.smantlement DAC Derived Air Concentration DAD Digital Alarming Dosimeter DAW Dry Active Waste DBE Design Basis Earthquake DECON 1mmediate Decontamination / Dismantlement Decommissioning Option DOE Department of fnergy DOT Department of Transportation DPM Disintegrations per minute DTS Decommissioning Technical Specifications EAB Exclusion Area Boundary EAL Emergency Action level ECP Executive Command Post EFPD Effective full Power Days E0C End of (fuel) Cycle EOF Emergency Operations facility EPA Environmental Protection Agency EPRI Electric Power Research Institute EPZ Emergency P1anning-Zone ERf Emergency Response facili+y ESW Equipment Storage Wells FCP forward C'/,nmand Post (EOF) X
6/28/91 PROPOSED DECOMMISSIONING PLAN TABLE Of CONTENTS fHM fuel Handlino Machine FNAL Fermi NatMg Atomic Laboratory " FSV fort St. Vrain fsW fuel Storage Wells GA General Atomics GET General Employee Training GM Geiger Mueller GTCC Greater Than Class 'C' (Radioaci ,e; Vu+e HEPA High Efficiency Particulate Air Li'ter) HLRW High Level Radioactive Waste HIWR High level Waste Repository We Hyper-Pure Germanium HSF Hot Service facility HTGR High Temperature Gas-Cooled Reactor HVAC Heating, Ventilation and Air Conditioning 100 Idaho Operations Office INEL Idaho National Engineering Laboratories INP0 Institute of Nuclear Power Operations IPEEE Individual Plant Examination of External Events IPP Independent Power Producer ISFSI Independent Spent fuel Storage Installation K1 t'otassium Iodide (tablets) LANL Los Alamos National Laboratory LLD Lower Limit of Detection LLRW Low-Level Radioactive Waste LSA low Specific Activity MCRB Metal Clad (Reflector) Block MDA Minimum Detectable Activity MicroR ;E(-6) Rem MVDS Modular Vault Dry Storage (System) NAVLAP National Voluntary Laboratory Accreditation Program NDE Nondestructive Examination NFS Nuclear fuel Services NFSC Nuclear facility Safety Committee ! NIOSH National Institute of Occupational Safety and Health ! NIST National Institute of Standards and Technology l NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System l OCC Office of Consumer Counsel ! ORE Occupational Radiation Exposure OSHA Occupational Safety and Health Administration
.PCC Personnel Control Center pCi Pico Curie (1 E-l? Curies)
PCP orocess Control Program E PCRV Prestressed Concrete Reactor Vessel PDP Proposed Decommissioning Plan xi
PROPOSED DECOMMISSIONING PLAN 6/28/91 TABLE OF CONTENTS POPC Plant Operations Review Committee PURPA Public Utility Regulatory Policies Act PSC Public Service Company of Colorado QA Quality Assurance QC Quality Control R/B- Release to Birth Rate RCA Radiologically Controlled Area l RCD Region Constraint Device l RCRA Resource Conservation and Recovery Act REM Roentgen Equivalent Man (Radiation Measure) : REMP Radiological- Environmental Monitoring Program ) RIV Reactor Isolation Valve S/G Steam Generator
- SAFSTOR Delayed Decontamination / Dismantlement Decummissioning l
Option SAR Safety Analysis Report SE0C State Emergency Operations Center SFSC Spent fuel Shipping Cask SR9 Self Reading Dosimeter TEDE Total Effective Dose Equivalent TLD Thermoluminescent Dosimeter TRU Transuranic Waste TS Technical Specifications TSCA Toxic-Substances Control Act TSC Technical Sepport Center UFSAR Updated FSAR UMTRAP Uranium Mill lailings Remedial Actions Project WBS Work-Breakdown Structure WITS Waste Inventory Tracking System WSEG Westinghouse Scientific Ecology Group xii
6/28/91 PROPOSED DECOMMISSIONING PLAN TABLE OF CONTENTS COMMONLY REFERENCED ISOTOPES AND ELEMENTS Boron B Calcium Ca-41, Ca-45 Carbon C-14 Cesium Cs-134, Cs-137 Cobalt Co-60 Dysprosium Dy Europium Eu 152, Eu-154 Fluorine F Germanium Ge Helium He Iodine I-129, 1-131 1ron Fe-55, Fe-59 Krypton Kr-90 Lithium Li-6, Li 1 Manganese Mn-54 Nickel Ni-63, Ni-59 Niob'.um Nb-94 Silver Ag-110m Strontium Sr-90 Tellerium Te-127m Tritium H-3 Xenon Xe-137 xiii
PROPOSED DECOMMISS10nlNG PLAN 6/28/91 TABLE OF CONTENTS INTENTIONALLY LEFT BLANK I xiv
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 1 SECTION 1
SUMMARY
OF PLAN
1.1 DESCRIPTION
OF DECOMMISSIONING PLAN AND DECOMMISSIONING ALTERNATIVE 1.1.1 IBinduction By letter to the Nuclear Regulatory Commission (NRC) dated December 5, 1988 (Ref. 1), Public Service Company of Colorado (PSC) notified the NRC that " based on economic considerations associated with the ongoing operating costs of Fort St. Vrain, PSC has determined that it will be necessary to terminate Fort St. Vrain operations early." At that time, PSC began decommissioning planning to support premature decommissioning, resulting in submittal of the Preliminary Decommissioning Plan to the NRC on June 30, 1989. (Ref. 2) This Prop < .ad Decommissioning Plan is submitted by PSC in accordance with the .iquirement of 10 CFR 50.82(a), which requires submittal of the Proposed Decommissioning Plan "within two years fo77owing permanent cessation of operations." PSC previously provided a target date of October 31, 1990, for submittal of the Proposed Decommissioning Plan. The Proposed Decommissioning Plan represents a departure from PSC's Preliminary Decommissioning Pl an (Ref. 2) in that, after consideration of financial risks, regulatory environment, and uncertainty of other issues, PSC has selected the DECON alternative for immediate dismantlement and decommissioning of Fort St. Vrain. Through a competitive bid process, PSC has selected a team headed by the Westingho Electric Corporation to carry out the decommissioning of Fort St. Vrain on a fixed price basis. Coincident with decommissioning, the Fort St. Vrain plant may be converted to a fossil-fueled facility (See Section 5.5). 1.1.2 B3ckaround Fort St. Vrain was shutdown on August 18, 1S09. On August 29, 1989, the PSC Board of Directors reviewed and confirmed the Executive Management decision that Fort St. Vrain would not be restarted, and that PSC would pursue the decommissioning of Fort St. Vrain. The decision to permanently shut down and decommission Fort St. Vrain was based on related technical and financial considerations. Problems were identified with the control rod drive assemblies and 1.1-1 1
PROPOSED DECOMMISSIONING. PLAN 6/28/91 SECTION 1 the steam generator steam ring headers that presented significant technicL1 obsh les which could be overcome, but at significant cost in dollars and time to PSC. Additionally, due to the uniqueness of the one of-a-kind High Temperature Gas Cooled Reactor (HTGR) fuel cycle, the cost to purchase new fuel was prohibitive, This, in conjunction with low plant availability and correspondingly high operating costs, made continued operation of Fort St. Vrain imprudent. Ceupled with these technical and fuel cycle considerations, Fort St. Vrain had previously been removed from the rate base as a result of a 1986 Settlement Agreement between PSC, the Colorado Public Utilities Commission (CPUC), the Office of Consumer Counsel (OCC) and other parties. With - the exception of limited funds to be collected for decommissioning, the removal of Fort St. Vrain from the regulatory rate base left PSC shareholders responsible for further operating and decommissioning costs of Fort St. Vrain. L 1.1.3 Contents of the Proposed Decommissionino Plan The Proposed Decommissioning Plan has been prepared to be responsive to the requirements of 10 CFR 50.82(b) and the guidance of Draft Regulatory ' Guide DG-1005 " Standard Format and Content for , Decommissioning Plans for Nuclear Reactors" (Ref. 3). The following L is a brief summary of the sections contained within this plan. Section Descriotion t 1 " Summary of Plan" provides a brief description of - the proposed plan 'and background information related to the decision to -decommission Fort - St. . Vrain. Information is provided to describe .the major activities involved in the dismantlement and decommissioning of Fort St. Vrain, and the j projected project schedule. The cost to decommission Fort St. Vrain is identified, as well as status of the
. availability of funding.. Details are provided in Section 1 1.4 on implementation and administration of the proposed l plan. Section 1.5 describes the controls which will be l
effective during the transition period prior to approval of the Proposed Decommissioning Plan. 2 " Choice of Decommissioning Alternative and Description of Activities" identifies the- selected decemmissioning alternative. Section 2.2 provides a description of Fort St. Vrain and identifies major site factors, and identifies l 1.1-2
6/28/91 PROPOSED DECOMMISS10hlNG PLAN SECTION 1 Section Description 2 contamina W or activated structures and components which (cont.) will be -noved during decommissioning. The- major decommissioning activities, schedule and exposure estimates are provided in Section 2.3. Organizational structures are provided_ for the PSC organization (Section 2.4) and the selected contractor (the Westinghouse team, Section 2.6). Decommissioning training requirements are identified in Section 2.5. 3 "Pt otection of Occupational and Public Health and Safety" describes the "as-is" radiological status of the Fort St. Vrain facility (Section 3.1). The decommissioning radiation protection organization is described in Section 3.2, and proposed methods of managing radioactive waste, including offsite transportation and disposal, are discussed in Section 3.3. The analysis of postulated bounding decommissioning accidents is provided in Section 3.4. 4 " Final Radiation Survey Plan" provides the purpose, criteria, and methodology that will be used to formulate the final radiation survey plan, including instrumentation, documentation and quality assurance requirements, and eventual site closure. 5 " Decommissioning fixed Price Contract and Funding Pl an" provides a description of the decommissioning fixed price contract, major assumptions and bases used to derive the
- decommissioning cost, and status of decommissioning funding.
Provisions are also identified for -updating -both the decommissioning cost and the funding plan.
- 6. " Decommissioning Technical and Environmental Specifications" provides the methodology and philosophy that will be'used to develop the decommissioning technical specifications. These specifications will be submitted to the NRC . in the near future.
( 7 " Decommissioning Quality Assurance Plan" provides the QA ! plan which will be effective during decommissioning. l 8 " Decommissioning Access Control Plan" identifies those access control requirements to be administered during the decommissioning process once all spent fuel has been removed 1.1-3
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 1 Section Descriotion M 8 from the Protected Area. T).i s access control plan will (cont.) replace the existing physical security plan during the decommissioning period. 9 " Decommissioning Emergency Response Pl an" provides an overview of the basis for emergency response during the decommission hg process. Accidents evaluated in Section 3.4 will be used as the basis for this plan, which will be submitted to the NRC at a future date. 10 "Decommissicning Fire Protection Plan" provides an overview of fire protection provisions which will remain in effect during decommissioning. Accidents evaluated in Section 3.4 are the basis for this section also, and a separate Decommissioning Fire Protection Plan will be submitted to the NRC at a future date. Appendix 1, " Westinghouse Team Scope of Work", provides a detailed description of the proposed Westinghouse team decommissioning and dismantlement activities. Appendix II, " Fort St. Vrain Activation Analys i s" , provides the results of the analysis to identify activation levels and isotopes for Fort St. Vrain components. 1.1-4
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 1 1.2 MAJOR TASKS, SCHEDULES AND ACTIVITIES 1 1 Description of Major Activities The major dismantlement and decontamination activities to be performed during decommissioning are described in detail in Section 2.3. The decommissioning project is divided into three major work areas:
- 1. Decontamination and dismantlement of the PCRV.
- 2. Decontamination and dismantlement of the contaminated balance of plant (B0P) systems.
- 3. Site cleanup and final site radiation survey.
Site cleanup is described in Section 2.3 and the final site radiation survey is described in Section 4. 1.2.2 Final ~ Release Criteria The release of the site, facilities and materials will be based on proper application of release criteria for surface contamination, soil / water concentrations and exposure rates. Final site release criteria are fully identified in Section 4.2 of this plan. 1.2.3 Decontamination and Dismantlement of the PCRV The major decommissioning task is the dismantlemerJ. and decontamination of the radioactive portions of the Prestressed Concrete Reactor Vessel (PCRV). Section 2.3 provides a comprehensive description of the steps necessary to dismantle and decontaminate the PCRV. PCRV dismantlement activities will begin only after all irradiated fuel has been removed from the Reactor Building. PSC and the Westinghouse team have evaluated technical options available for dismantling radioactive portions of the PCRV, and a decision has been made that the best technical approach is to flood the PCRV with water, and perform the majority of dismantlement activities submerged. This will allow the most direct access to highly radioactive portions of the PCRV, while affording the maximum shielding benefit. l The description and sequence of major activities associated with l PCRV dismantlement is described in Section 2.3. o l 1.2-1
PROPOSED DECOMMISSIONING PLAN- 6/28/91 SECTION 1 1.2.4 Decontamination and Dismantlement of Contaminated Balance of Plant Systems for the purposes of this Proposed Decommissioning Plan, balance of plant systems refer to those contaminated or potentially contaminated plant systems outside the PCRV. Decontamination and dismantlement of contaminated or potentially contaminated balance of plant systems will be performed by one of the fcMowing approaches: (1) decontamination in place, (2) removal and decontamination, cr (3) removal and disposal as radioactive waste. Systems which are contaminated or potentially contaminated above releas"* Io limits requiring decontamination and dismantlement are described n, faction 2.3. 1.2.5 Schedule for Decommissionina Activities - The schedule for decommissioning activities is provided in Section 2.3.5 and Figure 2.3-15. The following is a brief description of the two phases of the Fort St. Vrain Decommissioning _ Project: Phase i Decommissionina Plannina Phase, with an estimated duration of 18 months, consists of initial site characterization, preparation of work scope planninh work specifications and procedures, and equipment and material staging. There will be N0 physical decommissioning activities performed as part of this planning phase, although some component removal and disposal 3ctivities may occur prior to commencement of Phase 11-(described below) as described in Section 1.5 of this plan. Phase 11 Decontamination and- Dismantlement Phase, with an estimated duration of 39 months. Actual dismantlement, decontamination, and physical decommissioning activities - will occur as part- of this phase. The actual physical decommissioning activities are scheduled to commence after:(1) NRC approval of the Proposed Decommissioning Plan, and (2) removal of all irradiated fuel from the Reactor Building. It is important to note that Phase I and Phase Il activities are not conducted in series. These two phases have considerable overlap. Further detailed descriptions of the work scope to be performed in each project phase are provided in Appendix I of this plan. 1.2-2
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 1 Some component removal activities will be conducted prior to commsncement of 'the Deconta:nination and Dismantlement Phase, as described in Section 1.5. Decommissioning . of Fort St. Vrain, including site cleanup and final site radiation survey, is expected to be completed by April 1995, i I' l.2-3
4 J. .-p # J J-4,1 .&. 4-.-. :L- c4 A. f -dA-cJM-.A EwAa-- a-iJ*pi 15 +<4. M *A. _) --.M-+.a =A J 4%u.4
- _ Am PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 1 t
INTENTIONALLY LEFT BLANK i. l l-i L l' 1.2-4 i
6/28/91 PROPOSED DECOMMISSIONING PLAN 4 SECTION 1 1.3 FIXED PRICE AND AVAILABILITY OF FUNDS ' l.3.1 Deconnissionino Cost Through the -competitive bid process described in Section 5.1, PSC selected, from among four qualified bidders, a project team of Westinghouse and MK-Ferguson as its decommissioning contractor. The competitive bid submitted by the Westinghouse team, together with an estimate of PSC decommissioning costs, results in a total decommissioning cost of $137,129,000 based on the anticipated year of expenditure (inclusive of escalation). Of this amount, the Westinghouse team's firm fixed price is $100,460,000. PSC's costs, as overall project manager and licensing coordinator, are estimated I to be $36,669,000. Assumptions used as the basis for these costs ! are identified in Section 5.3. The proposed Westinghouse team Scope of Work is provided in Appendix 1. A detailed cost estimate that is responsive to the requirements of 10 CFR 50.82(b)(4) has been prepared to support the costs and was submitted to the NRC on June 6,- 1991 (Ref. 4) . 1.3.2 Deconmissionino Fundina Plan As of March 31, 1991, the Fort St. Vrain decommissioning trust fund balance was approximately $26.0 million. Under terms of the 1986 Settlement Agreement, funds in the amount of approximately $1.1 million remain to be collected from PSC customers by the end of i 1991. A final decision has not been made on the funding plan for Fort St. Vrain' decommissioning. Section 5.5 provides a discussion on possible decommissioning funding alternatives being pursued by PSC and commits PSC to notify the NRC of the projected funding plan once ongoing funding alternatives are finalized. ( 1.3-1 ; l
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 1 INTENTIONALLY LEFT BLANK 1.3-2
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 1 1.4 REGULATORY BASIS FOR ADMINISTRATION OF THE PROPOSED DECOMMISSIONING PLAN This Froposed Decommissioning Plan has becn prepared and submitted to be responsive to the requirements of 10 CFR 50.82 and the guidance of Draft Regulatory Guide DG-1005, " Standard Format and Content for Decommissioning Plans for Nuclear Reactors" (Ref 3). The Proposed Decommissioning Plan is intended to govern the entire Fort St. Vrain decommissioning effort, and is to be maintained current as described in this section, if and when the need for plan changes occur. This plan is to be a- key component af the licensing basis of Fort St. Vrain during decommissioning as described below.
-The following documents shall constitute the decommissioning licensing basis of Fort St. Vrain, effective following the removal of all irradiated fuel from the Fort St. Vrain Reactor Building and receipt of NRC approval to commence decommissioning:
- 1. This NRC-approved Proposed Decommissioning Plan including:
- a. Applicable NRC decommissioning regulations as identified in this plan, and
- b. NRC regulatory guidance applicable to the decommissioning of Fort St. Vrain as identified in this plan.
- 2. The NRC approved Fort St. Vrain 10 CFR 50 license and the Decommissioning Technical Specifications.
- 3. Licensing basis correspondence between the NRC and PSC related to the decommissioning of Fort St. Vra a.
This Proposed Decommissioning Plan, following its approval by the NRC and the removal of all irradiated fuel from the Fort St. 'Vrain Reactor Building, shall supersede and replace the. Fort St. Vrain Updated Final Safety Analysis Report (UFSAR, Ref. 5). Following the completion of defueling, the final revision of the Fort St. Vrain UFSAR then in effect shall be retained as an historical document only, and all of the operational descriptions and commitments therein shall be. superseded in their entirety by this Proposed Decommissioning Plan. Essential safety features and functions which will be relied upon during Acommissioning are described and included in this plan. For the purposes of the Fort St. Vrain plant decommissioning, the provisions of 10 CFR 50.59 and 10 CFR 50.71(e) shall apply to and be implemented by this Proposed Decommissioning Plan and the l Decommissioning Technical Specifications. Any Proposed Decommissioning Plan change or activity which involves an unreviewed l.4 1 l
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 1 safety question as defined in 10 CFR 50.59 or requires a change to the Decommissioning Technical Specifications shall require approval by the NRC prior to implementation. Any Proposed Decommissioning Plan changes or activities that do not involve an unreviewed safety question and do not require a Decommissioning Technical Specification change, as determined by performing a 10 CFR 50 !9 safety evaluation, may be implemented by PSC without prior NRC approval. An annual report shall be submitted to the NRC per the provisions of 10 CFR 50.59 describing all Proposed Decommistioning Plan changes made under the provisions of 10 CFR 50.59 and the results of the associated 10 CFR 50.59 safety evaluation. Likewise, Proposed Decommissioning Plan updates shall be submitted to the NRC at least annually per the provisions of 10 CFR 50.71(e), This annual Proposed Decommissioning Plan update shall be current as of six months prior to the submittal date. The following plans, which require NRC approval for decommissioning and constitute a part of this Proposed Deccmmissioning Plan, shall be administered under the applicable provisions of the regulations as described in following Proposed Decommissioning Plan (PDP) sections: ELAN APPLICABLE REQUIREMENTS /PDP SECTION Quality Assurance Plan (2) 10 CF" 50.54(a), 10 CFR 50 AppenC x 8, and 10 CFR 71 Subpart H as described in PDP Section 7
- Access Control Plan (2) PDP Section 8 Decommissioning Emergency Response PDP Section 9
! Plan (l) l Fire Protection Plan (l) 10 CFR 50.48(a) and 10 CFR 50 Appendix A Criterion (3), as described in PDP Section 10 l Final Radiation Survey Plan (2) 10 CFR 50.82(b)(3) as described in PDP Section 4 Decommissioning Funding Plan (l) 10 CFR 50.82(b)(4) as described in PDP Section S 1.4-2
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 1 Notes: (1) Plans which will be submitted to the NRC for approval separate from this Proposed Decommissioning Plan. (2) Plans which are included in this Proposed Decommissioning Plan for NRC review and approval, l l. l u-l' l.4-3
J- J. +- 3A.- - 'NJ 4 4 JA --u.A_a_ma.-- .amd 4%a, l PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 1 i l l l 4 INTENTIONALLY LEFT BLANK 1,4-4 l l . -
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 1 1.5 DECOMMISSIONING CONTROLS DURING THE TRANSITION PERIOD PRIOR TO APPROVAL OF THE PROPOSED DECOMMISSIONING PLAN 1.5.1 Int rodu_c_110B This section describes the decommissioning controls that will apply to plant closure activities during the transition to the Decommissioning Technical Specifications (DTS) and the Proposed Decommissioning Plan (PDP) controls. The decommissioning of Fort St. Vrain (FSV) involves many planning and preparatory activities that will be performed prior to approval of the PDP. The requirements and controls that govern these plant closure activities are contained largely in the operational Technical Specifications (TS) and in 10 CFR 50.59. The Fort St. Vrain 10 CFR 50 license includes controls in the TS which are an appendix to the license. The Administrative Controls in the TS will remain generally unchanged until the DTS (see Section 6.1) are approved and implemented. Pending NRC approval, the DTS may be implemented concurrent with the PDP approval or it may occur at a later time. Decommissioning activities, therefore, may have to be initiated under the then current TS controls. Decommissioning shall be considered to begin with the first physical activity to remove contaminated equipment from Fort St, Vrain, after all irradiated fuel has been removed from the Reactor Building and after NRC approval of the PDP. Activities performed prior to NRC approval of the PDP are considered plant closure activities, in preparation for decommissioning. In Reference 6, the NRC stated that a licensee must: (1) comply with the requirements of its operating license and the regulations applicable to whatever mode or condition the plant might be in at a given time; and D) refrain from taking any actions that would materially and dem)nstrably affect the methods or options available for decommissioning, or that would substantially increase the costs of decommissioning, prior to NRC approval of a decommissioning plan. Fort St. Vrain is permanently shut down, cooled down and depressurized. Under these plant conditions, PSC considers that performing plant closure activities is within existing licensee authority provided they do not require a change to the Fort St. Vrain Technical Specifications or 10 CFR 50 license, do not involve an unreviewed safety question as defined in 10 CFR 50.59, do not limit the choice of reasonable decommissioning alternatives (i.e., 1.5-1
FROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 1 SAFSTOR, DECON, or ENTOMB), and do not substantially increase the costs of FSV decommissioning. PSC will not consider resumed operation as an option during the review of contemplated component removal and disposal activities (e.g., region constraint devices and helium circulators). Other plant closure activities that are not within existing licensee authority will be submitted for NRC approval, prior to their accomplishment. PSC considers these actions to be fully in compliance with applicable regulations and license iequirements. 1.5.2 Component Remo_y_al Activities Prior to the initiation of actual decommissioning activities, PSC may complete the removal of numerous components from the PCRV, including the helium circulators, control rod drive and orifice assemblies (CRDOAs), metal clad reflector blocks (MCRBs) and the region constraint devices (RCDs). These activities are performed as pl ant closure activities, outside the scope of the Proposed Decommissioning Plan. Some of these components may be removed prior to the completion of defueling, if the components have no required nor useful function during any planned or postulated defueling or shutdown conditions. In addition, several component removal activities may be performed in the interest of technology transfer with the Department of Energy (D0E), including removal of a steam generator ring header and bimetallic weld sample (s), and removal of high temperature helium purification system components. Other plant closure activities beyond the scope of the Proposed Decommissioning Plan are evaluated against the following criteria:
- 1. If the activity requires a change to the Fort St. Vrain Technical Specifications or involves an unreviewed safety question, as determined by a safety evaluation performed in accordance with the provisions of 10 CFR 50.59, prior NRC approval must be obtained.
- 2. If the activity has an adverse environmental impact, in that it disturbs environs not previously disturbed during plant construction or operation, prior NRC approval must be obtained.
- 3. If the activity precludes any of the allowable decommissioning alternatives (SAFSTOR, DECON, or ENT0MB),
prior NRC approval must be obtained. I 1.5-2
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 1
- 4. If the activity involves any significant increase in the total radiation exposure required for decommissioning, to the extent that a revision to the Proposed Decommissioning Plan is required, prior NRC approval must be obtained.
1.5.3 Transition to Decommissionino Controls After all nuclear fuel has been removed from the Reactor Building, the controls on plant closure activities will not be needed to ensure the safety of the core, but they will be needed to minimize radiological exposure to workers and the public, and to protect against an unplanned release of radioacuvity to the environs. Fort St. Vrain will maintain its 10 CFR 50 license, and regulations such as 10 CFR 50.59 will still apply. The transition to decommissioning controls will be relatively smooth because many of the existing requirements will continue to apply, although various details may differ considerably. Facility modifications that involve a change to the PDP will be reviewed pursuant to 10 CFR 50.59. The DTS will include Administrative Controls, such as organizational requirements, a safety review committee, , rocedural requirements, record keeping requirements, and reporting requirements. Upon NRC approval of the PDP, decommissioning controls will be phased-in in a controlled manner, as follows:
- 1. Surveillances and preventive maintenance activities for equip :nt no longer required to be operable will be suspended.
- 2. New decommissioning design controls may be implemented which will incorporate revised requirements for 10 CFR 50.59 evaluations and configuration management.
- 3. Procedures that are no longer needed will be deleted or piaced in a category which requires no additional maintenance of the procedure.
- 4. Procedures that relate to radioactive effluent controls will be retained until the DTS are issued. At that time, they will be revised as necessary to reflect the requirements of the Off Site Dose Calculation Manual and the Process Control Program.
1.5-3
PROPOSED DECOMMISSIONING-PLAN 6/28/91 SECTION 1 5.- After approval of the DTS, implementing procedures will be revised accordingly. 1.5.4 Mobilization Activities In preparation for the actual start of decommissioning activities, it may be desirable to install certain equipment items such as material handling and water purification equipment prior to formal approval of the PDP. These mobilization activities may be performed while there is still nuclear fuel being removed from the Reactor Building, provided each , activity is evaluated for its impact on the defueling operation. Al so, any physical modifications to an existing Fort St. Vrain system (e.g., piping connections, power connections) will be treated in accordance with the then current Fort St. Vrain Quality Assurance Plan. 1 l l l- 1.5 4 l
5/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 1
1.6 REFERENCES
FOR SECTION 1
- 1. PSC Letter, R.O. Williams (PSC) to J. Calvo (NRC), dated Decerber 5, 1988;
Subject:
"Early Termination of Fort St. Vrain Operations",(P-88422).
- 2. PSC Letter, A.C. Crawford (PSC) to NRC, dated June 30, 1989;
Subject:
" Fort St. Vrain Preliminary Decommissioning Plan",
(P-89228).
- 3. NRC Regulatory Guide DG-1005 " Standard format and Content for Decommissioning Plans for Nuclear Reactors" (Draft for Comment), September,1989.
- 4. ' PSC letter, Crawford to 'Jeiss, dated June 6, 1991;
Subject:
" Fort St. Vrain Decommissioning Cost Estimate", (P-91198).
- 5. Fort St. Vrain Updated Final Safety Analysis Report, Rev. 8, dated-July 22, 1990.
- 6. NRC Memorandum and Order, CLI-90-08, dated October 17, 1990.
L l 1.6-1
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6/28/91 PROPOSED DECOMMISSIONING PJA SECTION 2 2.2 FACILITY DESCRIPTION 2.2.1 General Description Fort St. Vrain is a High Temperature Gas-Cooled Reactor (HTGR) owned and operated by PSC. Fort St. Vrain's location is approximately 35 miles north of Denver and three and one-half miles northwest of the town of Platteville in Weld County, Colorado. The site con!ists of 2798 acres owned by PSC. During the plant operation, approximately one mile square within the site area was designated as the exclusion area, and the licensee maintained complete control over this area. The completed facility is shown in Figure 2.2-1. The basic installation consists of a Reactor Building, a Turbine Building, cooling towers, and an electrical switchyard, 2.2.1.1 Reactor Buildina The Reactor Building (Figures 2.2-2 and 2.2-3) houses the
-prestressed concrete reactor vessel (PCRV), fuel handling area, fuel storage wells (FSWs), fuel shipment preparation facilities, decontamination and radioactive liquid and gas waste processing equipment, and most reactor plant process and service systems. The building is able to withstand wind loadings developed by a 100 mph wind or a tornado of 202 mph total horizontal wind velocity without exceeding yield stresses.
The Reactor Building ventilation exhaust filter system is designed to filter the Reactor Building atmosphere prior to release to the vent stack during both normal and most accident conditions during-decommissioning. The Reactor Building is maintained in a subatmospheric condition to ensure that all air leakage will be inward and to minimize unfiltered fission product release from' the building. The ventilation system was designed to maintain a l subatmospheric condition approximately 1/4-inch water gauge l negative. In actual practice, the Reactor Building pressure is normally 0.15 to 0.20 inches water gauge negative, _, depending on l building activities and ventilation system configuration. l l The PCRV and nuciear steam supply system (NSSS) ce located in the west partion of the Reactor Building. The east portion of the Reactor Building houses auxiliary and support systems and facilities such as the FSWs, the hot service facility (HSF), the equipment storage wells (ESWs), s'orage and laydown areas for various pieces l 2.2-1
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 Following defueling, the PCRV will contain the majority of the remaining radioactive materials in the Reactor Building. The Fort St. Vrain systems associated with the PCRV are as follows: System 11 PCRV and Internal Components System 12 Control Rod Drive and Orifice Assembly System 17& Reactor Reflector and Defueling Elements System 18 System 21 Helium Circulators System 22 Steam Generators System 23 Helium Purification System j i These systems make up the primary reactor vessel and internal core components located within the PCRV. These systems and components are discussed further in this section and in Section 2.3. 4 Portions of _ the PCRV concrete and rebar are expected to remain activated due_ to direct irradiation from the reactor core. Highly radioactive components will remain inside the PCRV until removed during PCRV decontamination and dismantlement. Physically, the 15-1/2 foct thick heads and the 9 foot thick concrete walls are constructed around a 3/4-inch thick low-carbon steel liner which forms the internal cavity. The liner is anchored to the concrete at frequent intervals. A core support floor (CSF) is provided within the PCRV in the form of a reinforced 5 foot thick concrete disk with a 3/4-inch carbon steel outer liner, supported by 12 steel core support floor columns from the bottom of the PCRV cavity. Longitudinal, circumferential and top and bottom crosshead prestressing tendons (448 total) are located in conduits embedded in the PCRV concre+e. Tendons are positioned both circumferentially and vertically along the PCRV side walls.- There are also tendons across the top and bottom heads in a criss-cross arrangement. The reactor core arrangement within the PCRV is shown in Figure
-2.2-7. The top layer of the core arrangement consisted of hexagonally shaped metal clad reflector blocks (MCRBs) with openings for 37 control rod pairs. The' MCRBs provided an inlet plenum for the reactor coolant to the active core. Region constraint' devices (RCDs) were located on top of the MCRBs and mechanically interlocked the top layer (not shown'on Figure 2.2-7). Hexagonal top reflector elements with coolant channels are located directly below the MCRBs and above the active core region.
2.2-3
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 The active core was divided into 37 regions and consisted of 1482 fuel elements. Individual fuel elements were hexagonal in cross section and aligned with the coolant channels from the reflector elements and MCRBs. During reactor defueling, the fuel elements are being replaced with defueling elements of identical shape and size. Hexagonal reflector elements are also located to the sides of and below the active core region. Many of the bottom reflector elements contain boronated graphite in Hastelloy cans. Radially outside of and immediately adjacent to the op, side and bottom hexagonal reflector elements are the large irregular-shaped side reflector blocks. Between the side reflector blocks and the core l'arrel are the boronated side reflector spacer blocks that contain boronated steel pins and were used for shielding. Immediately outboard of the core barrel is a helium interspace area. Outboard of this interspace area is an outer metal insulation cover plate, Kaowool (thermal) insulation, an inner metal insulation cover, another layer of Kaowool, and then the PCRV carbon steel liner. See figure 2.2-8 for :. general arrangement of the thermal barriers. Below the core region- containing the defueling elements, the CSF i will bear the weight of the defueling elements and reflectors through the core support posts and the core support blocks. The CSF also is the bottom termination point of the core barrel and has 12 penetrations for .the 12 steam generator modules. The CSF is -, supported from the bottom head of the PCRV with 12 core support I floor columns (See Figure 2,2-9). The lower plenum is below the CSF and houses the steam generator modules (12), circulator diffusers (4), circulators (4) the CSF support columns (12) and the lower floor. A number of instrumen and equipment penetrations and wells exist in the PCRV heads and sidewalls. 2.2.3 Balance of Plant Contcminated Components The systems id_entified below are considered to be the potentially contaminated balance of plant systems outside of the PCRV at Fort St. Vrain. Decontamination and dismantlement of these BOP systems are discussed in Section 2.3.4. System 13 h.a1 Handling Equipment System 14 Fuel Storage Facility System 16 Auxiliary Equipment 2 2-4
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 System 21 Helium Circulator Auxiliaries System 23 Helium Purification Auxiliaries System 24 Helium Storage System System 46 Reactor Plant Cooling Water System System 47 Purification Cooling Water System System 61 Decontamination System System 62 Radioactive Liquid Waste System Systers 63 Radioactive Gas Waste System System 72 Reactor Building Drain System System 73 Reactor Building Ventilation System System 93 Instrumentation and Controls System 15, fuel and reflector shipping equipment, consists primarily of the shipping casks, truck-trailers, spent fuel container, and cask lifting apparatus and is not a part of the decommissioning project. These equipment items will be retained under their separate 10 CFR 71 license or will be disposed of at some time in the future. A brief summary of the major components in each of the above balance of plant contaminated systems is as follows: 2.2.3.1 System 13 - Fuel Handlina Eauipment The fuel handling equipment that remains contaminated includes the fusi handling machine (FHM, Figure 2.2-10), five reactor isolation valves (Figure 2.2-11) and two refueling sleeves (Figure 2.2-12). 2.2.3.2 System 14 - Fuel Storaae Facility The fuel storage facility (See Figure 2.2-13) consists of nine fuel storage wells constructed of carbon steel liners suspended in concrete pits. 2.2.3.3 System 16 - Auxiliary Eauipmen _ The auxiliary equipment consists of the Auxiliary Transfer Cask (ATC, Figure 2.2-12), ten ESWs (Figure 2.2-14), the HSF (Figure 2.2-15), and three shielding adapters (Figure 2.2-16). The ATC is mnst commonly used to transfer the control rod drive assemblies, refueling sleeves and the shield plugs. The ten ESWs are carbon steel structures embedded in concrete used to store the control rod drive assemblies and the refueling sleeves. The HSF, constructed of concrete and steel shielding plates, consists of two 1 2.2-5
PROPOSED DECOMMISSIONING Pl.AN 6/28/91 SECTION 2 work areas used for inspection, repair, maintenance, testing and decontamination work. Figure 2.2-16 shows a general layout at the location of the various fuel handling and storage system components, and associated auxiliary equipment on the refueling floor. 2.2.3.4 Syltem 21 - Helium Circulator fux111 aries The auxiliary equipment for System 21 was used to provide -a su; ply l of high pressure water for the helium circulator bearing lubrication i
. and a supply of purified buffer helium to prevent in-leakage of bearing water into the primary coolant. The major equipment items include buffer helium recirculators, heat exchangers, filters, pumps, helium - dryers, chemical injection components, containment tanks,andcompressors(SeeFigure2.2-17). j 2.2.3.5 System 23 - Helium Purification Sys.tm The System 23 auxiliary equipment was used to assist in purification of the helium used as the primary reactor coolant. The major equipment iter.is include filters, heat exchangers, compressors, and dryers (See figure 2.2 18).
F.2.3.6 System 24 - Helium Storace System The primary purpose of the helium storage system was to provide for both storage and transfer of helium from the reactor vessel and the storage tanks. In addition, the nelium storage system was used in testing the control rod reserve shutdown system and for. various FHM purging operativ the primary equipment ' items include a helium transfer compres. c. e orage tanks, surge tank, oil adsorber, and high pressure heliv.. ,,upply tanks (See Figure 2.2-19). 2.2.3.7 System 46 - Reactor Plant Coolina Water System The reactor plant cooling water system (Figure 2.2-20) provides cooling water for process heat removal from all auxiliary equipment in- the reactor plant. Three loops are provided that form the PCRV circuit (liner cooling tubes), the PCRV auxiliary circuit (closed loop for various systems / components) and the service water circuit (open loop for various systems / components). The major equipment items include surge tanks, pumps, demineralizers, filters, heat exchangers, chemical injection (tank and pump) and recondenser chiller. 2.2 - -
6/28/91 PROPOSED 9ECOMMISS10NING PLAf4 SECTION 2 1.2.3.8 System 47 - purification Coolinn Water System The purification cooling water system (two loops) provides cooling
.;ater to the helium purification system heat exchangerr. lhe major components are pumps, expansion tanks, exchangers and associated piping (See Figure 2.2-21).
2.2.3.9 System 61 - Decontamination System The major equipment items iiiclude a water heater, a drying air heater, a filter, pumps,a solution tank and a chemical injection ' system (See Figure 2.2-22).
- 2.2.3.10 System 62 - Radioactive Liagid Waste Systs The major equipment items in this system include a waste sump (1000 gallon tank), pumps, filters, two 3000 gallon receiver tanks, two domineralizers and a 3000 gallon waste monitor tank (See figure 1 2.2-23).
2.2.3.11 System 63 - Radioactiye jas Waste System The major equipment items in this system include pre filters, filters, exhaust blowers, tanks (vacuum, surge, and drain), and compressors (See Figure 2.2 24). 2.2.3.12 System 72 - Reactor Build;no Drain System l The Reactor Building drain system collects the liquid effluent from l various equipment and piping drains for appropriate disposal. Thu l major equipment items include drain tanks, sump, pumps, piping and filters (See Figure 2.2 25). 2.2.3.13 System 73 - Reactor Buildina Ver;ilat1.pn System l The Reactor- Building HVAC system services various areas of the h Reactor Building with heated or cooled air. All ventilatOn air, , whether outdoor or recirculated, is filtered before dist, tbution. l In addition, the reactor plant HVAC system maintains building l- pressure differential control and collects radioactive leakage to ininimize exposure of personnel to airborne contamination. As shown in figure 2.2-26, this system consists of several air handling units l and filters. The only part of the system considered to contain possible contamina. tion is the reactor vent exhaust system. The reactor plant exhaust filters are composed of banks of . moisture 2.2-7
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 separators, HEPA filters and charcoal absorbers. Each bans contains 16 individual HEPA elements. 2.2.3.14 System 93 Instrumentation and Contr.gl lhe portions of the instrumentation and control system that are of interest i PCRV penetrations. These are orijitt,t thermocouple penetr .. g ,on's , t he . pr_ocess and moisture instrumentation, h110mcirculatorinstrumentation,andhaliumventp_iping1 21.4 Site Characteristics 2.2.4.1 Demoaraohv The population density in the rural areas surrounding the site is relatively low. The nearest resident is located approximately one half mile north of the Reactor Duilding, with the nearest town of Platteville located approximately 3-1/2 miles southeast. This is well outside the proposed EPZ of 100 meters from the Reactor Building. The population of Platteville, based on preliminary 1990 census figures, is 1515. The nearest population centers with a population over 25,000 are Greeley (60,399), Longmont (51,288), and , Loveland (37173), all based on preliminary 1990 census figures. 2.2.4.2 Geoaranhv and Land Use The site is located in Weld County, Colorado. The area surrounding i the site is shown in Figure 2.2 27 with reference circles of 10, 20 and 30 miles radii. The site is located in the South Platte River
- j. Valley, approximately thirty-five miles north of Denver, it is located in an agricultural area with gently rolling hills. Grade elevation at the plant is 4,790 feet. The foothills of the Rocky Mountains start to rise.about twenty miles west of the site, and the Continental Divide is prorainently identified by Long's Peak, located L forty miles directly west of the site.
The South Flatte River and St. Vrain Creek both pass through portions of the site. These two streams, which join near the northern tip of the site, are not large enough to be used for water transportation. l l The general area and land use. surrounding the site is predominantiy agricultural. Tne major farm- products include grain, feed corn, ! sugar beets, vegetables, beef cattle, sheep and turkeys. There is j also a limited amount of dairy farming in the area. l 2.2 8 i i 1
. . - . . . , ~ . _ , - . . _ . - _ . . . . - . - ___ . . . _ _ _ - - _ _ _ . . _ . - . .
6/28/91 PROPOS[D DECOMMISSIONING PLAN SECTION 2 The industrial facilities in the immediate area are primarily located in the town of Platteville. There are 14 oll/ gas wells within a one mile radius of the Reactor P'ilding on Company property. 2.2.4.3 Geoloav and Seinnol_qqr The geologic structure of the general area in which the site is located is shown in Figure 2.2 28. The area lies on the east-flank of the Colorado Front Range which is a complex 1y faulted anticlinal arch on which are superimposed numerous smaller folds and faults. The rocks of the core of the anticlinal arch are Precambrian crystallines, including gneiss, schists, and quartzites which have been intruded by granitic rocks that range in ago from Precambrian to Tertiary. On th'. east flank of the arch are Paleozoic and Mesozoic sedimentary rocks. The regional structure of this part of Colorado is characterized by sedimentary rocks dipping eastward into the Denver Basin. Along the mountain front the regional structural pattern is interrupted by relatively small, en echelon anticlines that plunge to the southeast. In addition to the fold axes, two groups of faults have been recognized. The most notable occurs along the mountain f ront and includes a series of faults extending in a generally northwest-southeast direction from the Precambrian into the Paleozoic-Mesozoic sediments. The second group of faults has been recognized primarily in coal mines, located generally east of Boulder. These faults have a northeast-southwest orientation. Both groups of faults are relatively high angle faults. The faults and the minor folds are related to the uplift of the front Range which began in Late Cretaceous and continued into the Tertiary. The original field examination and photo interpretation of the area surrounding the site location failed to indicate any evidence of recent movement along any of the known faults. -There is no known evidence of any recent seismic activity in the immediate area to have caused any subsequent movement. The subsoils at the site are St. Vrain Platte River alluvial sands and gravel overlying Pierre shale bedrock, Genera'ly, 3 to 8 feet of loose to very loose clean sands (with occasionai .ilty and clay lenses) are underlain by 30 to 35 feet of medium dense, fine alluvial sands. These sands are underlain with 4 to 11 feet of ' medium dense to dense, slightly clay, sandy gravel. Continuing under the gravel, hard to very hard interlayered sandstone and claystone bedrock is found at depth 46 to 51 fect. Free water was 2.2 9
POPOSED DECOMMISS10NIN3 PLAN 6/28/91 SECTION 2
- found at a depth of about 23 feet. Estimated contours of the surface of the oedrock &nd the free water level are shown in Figures '
2.2-29 and 2.2 30. The shallow loose s ar.ds are capable of supporting only icw foundation pressures the medium dense sand will support moderate foundation pressures, and the bedrock will support high foundation pressures. 2.2.4.4 Hydroloay The site location is between the South Platte River and St. Vrain Creek about two miles south of the confluence of these two streams. Surface water rights are owned in four ditches which traverse portions of the site area, in addition, nineteen shallow wells are located on the site area. Flow of ground water on the site is toward the alluvial deposits of both the South Platte River and St. Vrain Crock. The contours of the water table indicate that the flow of ground water is predominately toward the South Platte River Valley (Figure 2.2-30). M"ch of the ground water comes from the South Platte River and St. Vrain Creek, such that the water table changes with the flow rate (elevation) in the two streams. Total precipitation, mostly in the form of rain, in the South Platte Valley is small and contributes relatively little to the ground water. 2.2.4.4.1 Plant Water Sunolv When the plant was operating, cooling water for the plant wa, supplied by the main cooling- tower and the service water tower. Make-up water for the main cooling tower was obtained from water diverted from the South Platte River and St. Vrain Creek, and supplemented by water from a system of six shallow wells. Make-up water for the service water tower is supplied by the domestic water system, with back up from the shallow well system. Potable water and water for closed systems in the plant, such as the secondary coolant system, is supplied by the domestic water line, which is connected to a main of the local water district. The local water district is the Central _ Weld County Water District, whose source of supply i s Cclorado Big Thompson Project water from Carter Lake, which i s located Deut twenty miles west of the site. The arrangement or the various water supply systems is shown in Figure 2.2-31. 2.2-10 --r.-, gr e TT + =r - 4- - - ' * * " ^ ' + - ' -
- 6/28/91 PROPOSED DEC014MISS10NING PLAN <
SECT 10N 2 2.2.4.4.2 Elant Effluent , Liquid effluent from the plant is discharged primarily from either the plant building drains or the cooling tower blowdown line. Miscellaneous turbine plant drains such as floor drains, the Turbine Building sump, and yard drains, are normally directed to the South Platte River via the continuation of the Goosequill ditch to the farm pond. A diversion box is provided in the Turbine Building drain line so that effluent can normally be directed into the Goosequill ditch. Under abnormal conditions which prevent discharge via the Goosequill ditch, effluent is alternatively directed to the St. Vrain Creek via a slough. The reactor plant drair.c flow to a diversion box from which the flow can be directed to the South Platte River via the continuation of the Goosequiel ditch or to the St. Vrain Creek via a slough. Further downstream from the plant, the Goosequill irrigation ditch flows into the Jay Thomas irrigation ditch and the combined stream flows into a 25 acre farm pond. The overflow from the farm pond flows into the South Platte River close to its confluence with the St. Vrain Crack. The drainage path via the Goosequill ditch and the pond is normally used. Three lined evaporation ponds (total surface area of 3.6 acres) are present and were utilized to receive chemically treated effluent (primarily produced by periodic regeneration of plant domineralizers) while the plant was operating. Two ponds are located 'a -few hundred feet northeast of the plant building. The other pond is located south of the switchyard. Use of surface water downstream from the site is limited almost entirely to irrigation. A diagram of the major tributaries and irrigation ditches on the South Platte River between the gaging stations at Henderson and Kersey is shown on Figure 2.2-31._ The plant site is located just upstream of the junction with the St. Vrain Creek, adjacent to the Jay Thomas Ditch.
- Analyses for ' the reactor site were conducted on the amount of l diversion and stream flows of the nearby water ways. from these l original analyses, it was concluded that effluent from the plant l would be carried primarily by the South Platte R
- ver except during l the irrigation season with allowance for reservoir storage.
Effluent in irrigation water would enter ground water in the alluvium and would eventually be transported back into the strata 2.2-11 l
PROPOSED DECOMMISSIONING PLAT l 6/28/91 ' SECTION 2 bed of the South Platte River. There have been no signifiant changes in the waterway flows or diversions to require new analyses. ! The sources of public water supplies within thirty miles of the site are given in Table 2.2-1. There are two towns downstream within this radius that presently obtain part or all of their water from wells in the alluvium of the South Platte River: Gilcrest and LaSalle. It has been common practice for farmers to obtain domestic water from shallow wells in the alluvium. Many of the who formerly used shallow wells as their source of domestic A now obtain water from the Central Neld County Water District. Ti iame district is the source of dorr.astic water for the plant. 2.2.4.5 Meteoroloav 2.2.4.5.1 General Climate The general clirrate around the fort St. Vrain reactor site is typical of the Colorado eastern-slope plains region. In this s al-arid region the precipitation averages 10 to 15 inches a year, mL 'y from thunderstorms in late spring and summer. The annual free water surface evaporation rate is about 45 inches per year (Ref. 2). The wind records show no dominant direction, although winds out of the north by northeast segment do occur with the greatest frequency, The winds are generally light (10 mph), with higher velocities occurring during various atmospheric disturbances. The weather is generally mild. Most seasons are characterized by low humidity and sunny days, with occasional, short-lived storm. bringing precipitation into the area. Relative humidity averages about 40 percent during the day and 65 percent -at night. Thermal radiation losses resulting from lack of cloud cover provide considerable variation in temperature from night to day. Although snowfall may be significant, the snow cover is usually melted in a few days.
' 2.2.4.5.2 Severe Weather l
l Tabulated below are temperature and precipitation records for three cities within 20 miles of Fort St. Vrain (see Figure 2.2-27). The recording periods were 1973-1988 (Brighton), 1931-1988 (Longmont), l rd 1967-1988 (Greeley). I 2.2-12
6/28/91 PROPOSED DECOMMISSIONING PtAN SECTION 2 i Briohton 1.ongmont Greelev ' . Max. Temp.(degreesF) 101 106 103 Min. Temp. (degrees F) 23 -36 -25 Max. Precip. Day (in.) 2.73 4.04 3.20 Max Snowfall - Month (in.) 22.1 32.1 37.3 l Based on information extracted from archived weather data collected from Fort St. Vrain's 60 meter meteorological tower for the period 1986 through 1989, the following weather extremos were observed: Maximum Temperature - 104 degrees F Minimum Temperature - 26 degrees F Maximum Wind Velocity - 48 mph at wind direction 6.5 degrees (NNE) Seasonally, winds tend to be strongest in the late winter and spring, the season with high chinook frequency, and again in the , summer, when thunderstorms occur frequently. Strong winds, especially under chinook conditions, have been observed on various occasions in eastern Colorado. The chinook winds are strongest immediately to the east of the mountain ridge and diminish rapidly over the plains with increasing distance from '- the mountains. The measurement records at the site from July 1986 to December 1989 3' reveal a prevalence of northerly and southerly winds caused by the shallow depress'in of the St. Vrain Creek and the South Platte River and by the proximity of the Rocky Mountains. The meteorological data for this period for the wind speed and duration and frequency . of distribution is contained in Tables 2.2-2 and 2.2-3, respectively. Northeastern Colorado has moderate thunderstorm activity. The region near fort St. Vrain averages 50 days / year in which thunder and lightning occur. The majority of these thunderstorms are present from late spring through the summer. The Fort St. Vrain site is located in a region that typically experiences 5 tornadoes per year per 10,000 square miles. The peak 2.2-13
l l 1 PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 tornado activity occurs in the month of June, According to the . National Weather Service, 117 tornadoes occurred in Weld County during the period 1950 1987. l l 2.2-14
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6/28/91 PROPOSEU DECOMMISSIONING PLAN SECTION 2 2.3 DECOMMISSIONING ACTIVlil[S PLANNING AND EXPOSURE ESTIMATf.S 2.3.1 IDitoductiou Decommissioning of fort St. Vrain includes the dismantlement, , decontamination and disposal of radioactively contaminated or potentially contaminated material and components within the PCRV, and in contaminated or potentially contaminated balance of plant ! systems, and on the remaining site, followed by the final radiation l survey. Some of the activities described in this section will be performed prior to approval of the Proposed Decommissioning Plan, and are considered plant closure activities in preparation for decommissioning. Section 2.2 provided a description of the facility and site characteristics. The activated and contaminated portions of fort St. Vrain which will be decontaminated, dismantled and removed during the decommissioning process are identified in-Sections 2.2, 2.3 and 3.1. The specific tasks to be performed to , accomplish this -goal are discussed in this section. Although personnel conducting the dismantling activities will be exposed to radiation above background levels, the dismantling and decontamination activities have been developed to limit exposure to and control radioactiva material in order to maintain occupational doses as low as reasonably achievable (ALARA). Exposure estimates to accomplish the individual tasks and overall project are also provided. To - accomplish the decommissioning of Fort St. Vrain, substantial portions of the existing plant will be dismantled and removed. However, Reactor and Turbine Building - components and structures which are not radioactive above releasable limits will remain. The decommissioning project is' divided into three major work areas: -
- l. Decontamination and dismantlement of the PCRV.
.2. Decontamination and dismantlement of the contaminated or '
potentially contaminated balance of plant systems.
- 3. Site cleanup and final site radiation survey, Site cleanup involves pre- and post-decommissioning surveys of the site, and the radiological decontamination necessary to meet the regulatory guidelines to allow release for unrestricted use. These activities are discussed in detail in Section 4 and are not addressed in this section.
2.3-1 _ _._ _ _ . -. _. _ _ _ . _ _ _ . _ _ _ _ _ . - , _4
1 PROPOSED DCCOMMISS10NING PLAN 6/28/91 : SECTION 2 ECRV Decontaminntion_anLDiagnLiment Attiviun The following are the major activities involved in dismantling and removing the radioactive portions of the PCRV. These activities will be discussed in further detail in the following Section 2.3.3: '
!. Initial PCRV Preparation
- 2. Removal of the IIelium Circulators
- 3. Stum Generator Disassembly
- a. Initial Preparation
- b. Removal of Steam Generator Secondary Assembly
- 4. Removal of Activated Components using the ATC and film
- 5. Detensioning and Removal of Pretensioned Tendons
- 6. Flooding of the PCRV
- 7. PCRV Top Head Concrete and Liner Removal
- 8. Dismantling PCRV Core Components
- 9. Removing the Core Barrel
- 10. Removal of the Core Support floor
- 11. Disassembling the PCRV Lower /lenum
- 12. Final Dismantling, Decontaminat '.on, and Cleanup Activities A technical evaluation is provided in Section 2.3.2 which provides the basis for the technical approach selected to decontaminate and dismantle the PCRV. A brief description is also provided to identify various techniques which were considered for removal of the
.PCRV activated concrete.
Balance of Plant System Decontamination and_ Dismantlement Activitiel The balance of plant systems that are contaminated or potentially contaminated above releasable limits and will require decontamination or dismantlement are identified in Section 2.2.3. Work activities associated with these systems are discussed in paragraph 2.3.4 of this section. 2.3.2 Iechnical Anoroach Selection 2.3.2.1 Options Considered for Remon1 of the PCRV Key elements of the decommissioning plan include the techniques to be used to remove the internal components from the PCRV and to remove the activated concrete from the PCRV structure. This technical approach is based on filling the PCRV with water for shielding while internal components are being removed and using diamond wire cutting to remove the activated concrete from the PCRV structuro. These methods provide the decommissioning project with 2.3-2
l l , 6/28/91 PROPOSED DEC0ftilSS10NING PLAN SECTION 2 the optimum schedule, cost ALApt, risk, and safety considerations for decommissioning the PCRV. A detailed description of the PCRV fisassembly techniques and the basis for selecting them are described below. Two basic methods to disassemble the PCRV were considered: (1) in air (dry) disassembly, and (2) filling the PCRV with water to provide shielding. Two possible methods of in-air dismantlement were also evaluated, considering f actors of ALARA, safety risks, , schedule and cost. The two in air methods evaluated were fully i remote disassembly through the refueling penetrations in the top head, and partially remote disassembly from a massive shielded work platform with the top head removed. The following paragraphs provide an evaluation of each method, discussion of advantages and disadvantages, and a determination of its acceptability. 2.3.2.1.1 fal)v Remote. In-Air Disassembiv; The fully remote, in-air approach to the PCRV disassembly relied upon the extensive use of compicx remote tooling and resultant limited view of dismantlement operations, which would produce less than predictable results. Although use of remote operations would potentially result in the best ALARA and safety records, all activities would be performed with highly specialized robots. Therefore, the risk of failure or project delays would be greater due to potential L,reakdowns or delays, lack of reliable backup techniques, and lack of adequate contingency plans. Design, fabrication and testing of specialized robotics would also have to occur in a relatively short period of time, which could cause unnecessary delays it, the project schedule.
~
Addit. anally, removal of the CSF would be extremely difficult, since it is too massive (270 tons) for practical remote removal. 2.3.2.1.2 Partially Remote. la-Air Din ssembiv: Partially remote in air disassembly of the PCRV relied upon a massive shleided work platform that would be required to protect workers from radiation- exposure during disassembly. Access ports would be required- in this platfore through which hand-held, pole-type tools could be inserted to perform the disassembly when the platform is properly indexed over the work location. Using this approach, radiation exposure would be increased because of the extended stay times resulting from restricted tool access. Removal i of the top head and the top of the PCRV liner for installation of the work platform would be difficult because of high radiation levels and would probably require remote operations. 2.3-3
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 2.3.1.2.3 Floodina the PCRV: 1he final approach evaluated was to flood the PCRV cavity with water. This selected approach will provide optimum shiciding and contamination control and will allow the PCRV disassembly to be completed with optimum balance of schedule, cost, ALARA exposure and minimum risks. Additionally, there is an inherent added measure of safety due to the passive nature of the water for shielding and contamination control. Dismantling operations are greatly l simplified by "line of sight" manipulations as a result of direct ; viewing of the entire cavity. ' 2.3.1.2.4 [pnclusion: The evaluation of the two " dry" (in air) approaches against the
" wet" approach for PCRV disassembly favors filling the PCRV with water for shielding during disassembly. Additionally, it is noted that the " dry" techniques are not completely dry, since large volumes of water are required for any abrasive process used to cut the activated concrete inside the PCRV. Therefore, water would be !
introduced into the PCRV in each of the " dry" dismantlement options I considered. 2.3.2.2 lechniaues Considered for Removal of PCRV Activated Concrete Diamond wire cutting and abrasive water-jet cutting were evaluated for removing activated concrete from the PCRV walls. Diamond wire cutting was chosen as the method for cutting most of the concrote into sections because this proven- technology lends itself well to the PCRV concrete removal activities. Abrasive water-jet cutting was determined to be feasible for much of the concrete cutting but has been minimized to limit the production of contaminated abrasive waste and because. of related ALARA considerations.- The abrasive water-jet is presently being considered for one application, cutting of the CSF. The following techniques were also evaluated and were determined to be less desirable for the following reasons: (1) Expanding grout and explosives could be used to break apart the PCRV concrete, but were less desirable because of the heavy reinforcement of the concrete and the presence of the PCRV liner on the face of the concrete; (2) Thermal techniques were evaluated but were less desirable due to tool positioning difficulties, which could. cause cost and schedule concerns; (3) Mechanical irpact was evaluated but 2.3-4
l 6/28/91 PROPOSED DEC0hMISS10NING PLAN SECTION 2 were less desirable due to structural considerations (with the exception of removal of portions of the lowest concrete in the top head). 2.3.3 PCRV Dismantlement and DecontaminaLt10.0 2.3.3.1 Overview of PCRV Dismantlement Activities The major decommissioning task is the dismantlement and decontamination of the radioactive portions of the PCRV. A description of the PCRV is provided in Section 2.2 and illustrated on figure 2.2-6. It should be noted that the steps identified in the following paragraphs represent preliminary planning and may change during the detailed engineering and work development that will occur during the planning phase. This section provides a description of the expected steps necessary to dismantle and decontaminate the PCRV. Initial dismantlemeat of the PCRV will include removal of selected PCRV internal components and removal. of portions of the steam generators. The selected internal PCRV components will be removed from the upper portion of the PCRV using the fuel handling machine (FHM) and Auxiliary , Transfer Cask (ATC). These components may incit.de the 37 control rod metal clad reflector blocks (HCRBs), 270 non cont'o1 rod l hexagonal MCRBs, and certain helium purification components. '
-Simultaneously, the non contaminated portion of the steam generators (also called the steam generator secondary assemblies) will be removed f rom the lower portion of the PCRV to provide access for detachment of the contaminated steam generator primary assemblies (See figure 2.2 6).
To facilitate the removal of the remaining reactor core componants, the reactor cavity will be flooded with water. As discussed in Section 2.3.2, flooding the PCRV will provide shielding for the workers associated with PCRV dismantlement activities. After the steam generator secondary assemblies are removed from the bnttom of the PCRV, the PCRV bottom head and side wall penetrations will be scaled, a water cleanup and clarification system will be connected, and the PCRV will be flooded. To gain entry to the PCRV cavity, a plug of concrete will be removed from the top of the PCRV. ~ Selected PCRV prestressing tendons (See figure 2.2-6) will be either (1) detensioned an,d removed or, (2_1 detensioned and left in place. The top head plug will be cut into sections of appropriate size such that the weight and dimensions j will allcw the'i to be handled with the Reactor Building crane and 2.3-5
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 permit them to be moved out of the building. After the majority of the concrete has been removed from the PCRV top head, the 3/4 inch steel PCRV liner plate will be cut and removed with the remaining concrete, together with the top head liner insulation. A detailed discussion of this activity is provided in Section 2.3.3.7. Once access is gainad to the PCRV cavity, a work platform will be installed at the approximate elevation of the top of the PCRV where the liner and concrete have been removed. Working from this platform, workers will remove core components, including the remaining MCRBs, defueling elements, hexagonal reflector blocks, large side reflector blocks, side spacer blocks, core support blocks and core support pos,t s . This activity is described in Section 2.3.3.8. Once the core internals have been removed, the core barrel (a large carbon steel cylinder) will be removed by cutting it into pieces sized to fit in radwaste containers. (See Section 2.3.3.9) Following removal of the core barrel, the PCRV water level will be lowered and the CSF insulation removed, in preparation for removal . of the CSF. The CSF is a 29 foot diameter, 5 foot thick disk of reinforced concrete within a 3/4-inch steel casing weighing approximately 270 tons. The CSF will be detached from the twelve CSF columns and the twelve steam generator inlet ducts and lifted with a hydraulic jacking system to the PCRV top head region. The
, lacking system will then lower the CSF onto supports on the ledge in the cavity where the PCRV top head was removed. Once supported, the CSF will tw sectioned into,, segments small enough for handling by the Reactor Building crane. This activity is discussed in Section 2.3.3.10.
Once the CSF is removed, the PCRV lower plenum is exposed and the helium circulator diffusers and steam generator primary modules can be removed. These activities are discussed in Section 2.3.3.11. , The removal of the steam generator primary assemblics completes the removal of the major PCRV radioactive components. Remaining radioactive components include the activated " beltline concrete" around the reactor core region, the PCRV liner, liner insulation and insulation cover plates, and the PCRV lower floor with its supports. The activated beltline concrete is the PCRV region that was adjacent to the reactor core. It is estimated that this activated region is defined by a cylinder with an 18 to 24 inch wall thickness and a height of 40 feet. This section of PCRV sidewall will be removed by cutting and removing vertical segments. The activated liner plate, 2.3-6
6/28/91 PROPOSED DEC0!*11SS10 tilt 1G PL Att SEC110ti 2 insulation and cover plates will be removed with the ancrete. 1hese activities are discussed in Section 2.3.3.12. In the lower portion of the PCRV cavity (below the CSF), the insulation and insulation cover plates will be removed from the PCRV liner. The lower floor and all support members, insulation ard other components will be removed, and the exposed PCRV liner will be surveyed and decontaminated as appropriate. These activities are also discussed in Section 2.3.3.12. 2.3.3.2 initial PCRV Preparation Initial tasks to be completed in preparation for dismantling the PCRV will include acquiring tooling, setting up training mockups, installation of the PCRV water cicanup and clarification system, and craft personnel training in accordance with Section 2.6. Preparation activities include any modifications or revisions to existing f acilities and equipment and installation of new facilities and equipment that would be necessary for their use in supporting the decommissioning operations, fio major facility modifications are required that will affect the safety of the facility. Preliminary plans involve enlarging the refueling deck equipment hatch and truck bay door to allow passage of larger items. Plans also include re reeving of the Reactor Building crane to provide additional vertical travel which will allow the 170 ton main hook to travel from the refueling floor to ground level. This re-reeved configuration is consistent with the crane configuration used during original plant construction and is necessary to provide the lif ting capacity to lift the PCRV top head concrete block sections and other heavy lifts when components are rom.v d from within the PCRV. The need for extensive waste handling facilities in addition to those already present has been minimized by proper sequencing of the dismantlement activities and by proper management of the radioactive waste program, as described in Section 3.3. Off-site facilities will be utilized when necessary and practical for waste processing and final packaging. Proper task planning and sequencing will aid in minimizing accumulation of radioactive waste onsite. A self contained mobile laundry f acility to clean all contaminated protective clothing will be utilized. A water cleanup and clarification system, installed to maintain water purity in the flooded PCRV, will be discussed in Section 2.3.3.6. 2.3 7
i PP0POSID DECOMMISSIONINC PLAN 6/28/91 SECTION 2 Following helim circulator machine assembly removal, as identified in Secticn 1 S.2 (See figure 2.3'), each of the four nelium circulator P'.d penetrations will be scaled by installing a closure i fixture designed to withstand pressure when the PCRV is flooded with water. 2.3.3.3 Steam Genera;or Disassemb1v 2.3.3.3.1 Initial Stean Generator Disassembly Each of the twelve steam generators consists of a primary 6ssembly and a secondary assembly (figure 2.3 2). The primary assembly is located within the PCRV lower plenum and the secondary assembly is located beneath the primary assembly inside a PCRV bottom head steam generator penetration. The primary assembly is contaminated and the secondary assembly is not expected to be contaminated. However, in order to remove the prirary assembly, the secondary assembly must first be removed from beneath the PCRV. The removal of the insulation from the steam generator secondary side piping will be limited to the sections of feedwater, main steam, hot reheat, and ccid reheat piping that need to be severed for the steam generator secondary side removal. Prior to removal, i the insulation will be tested for asbestos content. If asbestos is present, appropriate controls will be impicmented for removal of the insulation, following the removal of the insulation, the main steam, feedwater, hot reheat and cold reheat piping will be cut which will allow the secondary side of the twelve (12) steam generators to be removed. 2.3.3.3.2 Removal of Steam Generator Secondary Assemb1v Removal of the steam generator secondary assemblies (See figure 2.3 2) will be accomplished in the reverse of the original i construction installation sequence. The steam generator secondary assemblies are expected to be free of contamination. The Marmon clamp (See figure 2.3-2) will be removed from the lower end of the steam generator secondary assembly. This will allow l withdrawal of the hot reheat piping from the steam generators. Because of the length of the hot reheat pipe, it will be severed into several sections as it is being withdrawn from the steam generators. The cold reheat pipe will then be severed at the threaded connection below the primary closure dome. Severing this connection remotely 2.3 8
6/28/91 PROPOSED DCCOMMISS10NING PLAN SECTION 2 or with a shielded man basket will minimize the exposure to workers while severing the cold reheat pipe. After the top of the cold reheat pipe has been cut, the lower reheat nozzle assemoly will be cut free of the steam generator secondary assembly at an elevation below the feedwater ring header. This will allow the withdrawal of the cold reheat pipe from the steam generator for disposal. After the cold reheat piping has been removed, the 40 feedwater, instrument, and steam tubes will be cut below the primary closure dome. The steam generator secondary assembly will then be rigged for lowering. The secondary closure weld will be cut and the steam generator secondary assembly will be lowered out of the PCRV penetration liner. The Rucker machine, which is a large turntable designed to handle heavy loads under the PCRV, will be used to handle the steam generator secondary assemblies in the reverse order of the installation operations, in order to detach the primary assembly from the penettation liner, the final step will be to unbolt the penetration liner from the flange joint at- the primary closure dome. Af ter unbolting, the primary assemblies will be resting on the liner penetration flange. The steam generator primary assembly is also stabilized by the steam generator shroud connection to the lower floor and the helium duct connection to the CSF. Each of the twelve steam generator primary assemblies will be detached from their respective penetrations in the above sequence and will then be removed through the top of the PCRV af ter the CSF is removed. This is discussed further in Section 2.3.3.11, When cutting operations have been completed, the interior of the penetration liner may be sprayed with a strippable coating to ease future decontamination operations. A new secondary closure plate will be welded in place to seal the penetration liner in preparation for flooding the PCRV. In parallel with the removal - of the steam generator secondary assemblies, the PCRV lower plenum (See Figure 2.3-3) will be entered through the PCRV bottom head access penetration after removal of the shield plug. A radiological survey of this area will be performed to determine radiation levels and major contributors in this area. Still photographs and video recordings will also be made to assist in mockup design and training for eventual dismantlement of the PCRV lower plenum. 2.3-9
l PROPOSED DECOMMISSIGNING PLAN 6/28/91 1 SECTION 2 2.3.3.4 Rem 2yal of Activated Components Usina the ATC and FHM Selected activated components will be removed from the PCRV using ' the ATC and the FHH. Use of this equipment will provide shielding while transferring highly radioactive components from the PCRV to i shipping casks with minimal personnel exposure. The 37 control rod : MCRBs and the 270 non control rod hexagonal MCRBs will be removed l from the PCRV by the fHM. Removal of certain components in the helium purification wells and penetrations, and placement of the refueling sleeve, will be performed by the ATC. As identified in Section 1.5.2, the RCDs, CRDOAs, and high temperature helium purification equipment may have been previously removed. , 2.3.3.5 Detensionina and Removal of Pretensioned TendQui 2.3.3.5.1 Tendon Removal Concurrent with operations discussed in Sections 2.3.3.2 through 2.3.3.4 is the detensioning of selected tendons in the PCRV. The PCRV_ has a total of 448 prestressing tendons made up_ of verticah circumferential, and top and bottom cross head tendons (see figure 2.2 6). The following identifies the number of tendons to be detensioned and also the number of tendons to be both detensioned and removed for each of the various tendon types. Number Of Number To No. of Tendons Type of Tendon Tendons De Detensioned ToBeRemoved Vertical 90 90 90 Circumferential 310 82_ 36 Top Cross Head 24 24 24 Bottom Cross Head 14 _Q _0 TOTALS: _ 448 196 150 Temporary . scaffolding will be used to facilitate tendon removal. Tendons will be detensioned by cutting individual tendon wires. 2.3.3.5.2 Vessel inteority The modified PCRV structure (see Figure 2.3-14j was evaluated for the loadings produced by the dead weight of the PCRV structure and components, the lifting operations of the CSF, and a design basis seismic event. Design response spectra from NRC Regulatory Guide 2.3-10
6/28/91 PROPOSED DECOMMISSIONING PLAN ; SECTIC1 2 1.60 " Design Response Spectra for Seismic Design of Nuclear Power Plants" (Ref. 3)_ and normaltred to the fort St. Vrain specific
" double earthquake" ground motions, together with damping values from Regulatory Guide 1.61 " Damping Values for Seismjc Design of.
Nuclear Power Plants" (Ref. 4), were inputs to the analysis along with the other structural loadings, in order for the analysis to be conservative, all PCRV upper half circumferential tendons were considered to be detensioned even though only the inner row of circumferential tendons .in the upper half of the PCRV beltline region need to be detensioned, along with all of the top head circumferential tendons. In summary, the concrete compressive and the reinforcing steel stresses of the modified PCRV are_within allowable limits and 1 provide adequate margin of safety for all loading conditions specified. The potential for cracking of c.oncrete in the modified top head and beltline regions has been reviewed and, considering the relatively low tensile . stress in a conservative number o_f reinforcing _ bars, the possibility of cracking due to tension in the concrete is considered to be extremely remote. ! 2.3.3.6 Floodino_Df the PCRV 2.3.3.6.1 Preparation for floodina the PCRV Once operations described in Sections 2.3.3.2 through 2.3.3.4 have been completed, activities may proceed to flood the PCRV. A network of PCRV liner cooling tubes (System 46) and the tendon tubes within the PCRV concrete wall creates a potential pathway for water leakage and the spread of contamination during the cutting of the PCRV concrote. To block these potential leak paths and prevent the spread of contamination, the liner cooling tubes and selected tendon tubes will be sealed with grout or or other suitable sealant. Before finoding the vessel, PCRV penetrations which are below the l PCRV waterline and have had their instrumentation removed (including i instrument penetration internal components and other items such as the thermocouples -routed through the core support blocks) will be sealed. These penetrations will be sealed by either one or a j combination of the - following: cutting and capphjust outside the ! PCRV or by installation of bolted and gasketted blind flanges. Where welding is utilized, all root passes will be liquid penetrant tested per applicable codes. A PCRV low point penetration will be sealed with a specially designed cover plate before the PCRV is flooded, This closure will provide suction and fill connections for 2.3 11
PROPOSED DECOMMISSIONING PLAN 6/20/91 SECTION 2 the water cleanup and clarification system described below, lite root pass for the welded connection will be_i_iguid penetrant tested. 2.3.3.6.2 Installation of the Water _Cleanuo and Clarification System After verifying that the steam generator and helium circulator penetrations are scaled, the water cleanup and clarification system will be installed. The water cicanup/ clarification system (See figures 2.3 4 and 2.3 5) will consist of two 50 percent trains of parallel equipment. Typically the system will be operated using both trains of equipment, but it can be operated at reduced capacity (one train only) to permit routine maintenance. Figure 2.3 5 is a pictorial view diagram of the planned system. Each train of equipment will consict of a coarse strainer designed to remove gross debris and to protect downstream equipment, a settling feed pump will discharge to_ a gravity set t1_ing tank. Sludge f rom the tank will be collected2 dewatered, and treated as _ radioactive waste. Discharge from_ the settling tank will supply a clarifying _ pump. A bank of _ filters will provide the degree of filtration necessary to ensure acceptable water clarity. Suitable valving- and cross connection between trains will enhance system ficxibility and availability. In addition to the capability for full-flow filtration of the PCRV water inventory, the system design will also include partial (side stream) demineralization for controlling dissolved solids. " feed and bleed" connections for adding clean makeup water and for removing contaminated water will also be provided to control tritium. (Tritium and liquid release is discussed further in Section 3.3). Chemical addition tanks are included in the design for chemistry and pH control, and to suppress biological growth. The system design will also include instrumentation, controls and sampling points. These will enable proper operation in monitoring the system and effectiveness of its components. The system will also be cross-connected to the existing Radioactive Liquid Waste System to allow discharge effluent to be processed by the existing system. The purified water will return to the top of the PCRV cavity by means of a distribution header designed to minimize local velocities and turbulence to maintain underwater visibility. The equipment will be appropriate for the radioactive nature of the process fluid. The recommendations of Regulatory Guide 1.143
'_'De s i gn Guidance for Radioactive Waste Management Systems,
( Structures2 and Components Installed in LJght-Water Cooled Nuclear l Power Plants" (Ref. 5), as well as ALARA considerations, will be 2.3 12 i L _ _
._.._ .._ _ _ _ ,_ - _ _ _ _ _ . - _ _ ___ _ __ _ m
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 used in the system design. EqJipment that can generate a high radiation field, such as filters, will be shielded and provided with remote handling capability. Equipment fluid drains and leakoffs will be collected, treated and disposed of as discussed in Section 3. 2.3.3.6.3. floodina the PCRV The PCRV will be flooded well before operations that require , shielding are scheduled, to allow time for the water chemistry to ' stabilize ar.d turbidity to be eliminated by filtration. The PCRV will be flooded to a level that radiological conditions necessitate to provide ample shielding when the PCRV top head concrete is removed. 2.3.3.7 PCRV Too Head Concrete and liner Removal It is planned that the PCRV top head will be cut using diamond wire techniques, and removed in several sections which can be handled with the re-reeved Reactor Building crane. These sections will be j cut so as to leave a thin horizontal layer of concrete above the ; PCRV liner. The remaining layer of activated concrete and liner l will be removed by breaking an annular portion of the concrete with - a mechanical breaker to expose the liner, then cutting the liner. This sequence is performed in this manner to prevent inadvertently breaching the PCRV liner and minimize exposure of equipment and personnel to radioactive material. The PCRV top head sequential cutting operations consist of the following major activities (the number and shapes of these sections may change _ based on detailed engineering evaluation during the planning phase):
- 1. Seal the top head penetrations to prevent debris from entering-the PCRV.
- 2. Set up the core drilling machines on the external wall of the PCRV to create five horizontal core drilled holes, i (figure 2.3 6).
- 3. Thread the diamond wire through the intersection points of the cored holes to make a loop to allow cutting of the concrete (figure 2.3-7).
- 4. Insert shims in the kerf of the diamond wire cut area to l prevent closing of the gap due to the weight of the concrete.
S. Make ten inclined core drilled holes to intersect with the horizontal cut kerf (figure 2.3 8). 2.3-13
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2
- 6. Make the five vertical sectioning cuts using the diamonti wire method (Fiqure 2.3 9).
- 7. Make the six vertical tapered back cuts using the diamond wire method.
- 8. Rig the sections for removal.
Removal of the ten sections of tophead concrete, of which the lower portions may be activated, will be accomplished utilizing the ; re reeved Reactor Building crane. This will leave a thin layer of activated concrete covering the PCRV liner. The ten concrete sections will be moved to a waste processing area for further sectioning, segregation and preparation for disposal. The diamond wire cutting system consists nf a wire with collars containing a diamond matrix, made to length for each individual cut, and a hydraulic pulley drive system to circulate the wire, lhe diamond wire is routed to envelop the cut area and then returned to a drive wheel on the drive system. The wheel rotates and pulls the wire through the cut areas. Ilydraulic cylinders control the tension of the wire. Once the cut is started, the tension is increased to optimize cutting efficiency. In order to remove the remaining concrete layer and liner plate with attached insulation, the following steps will be performed. Concrete will be removed providing a circular trough arot.nd the outer periphery of the reactor cavity (See figure 2.310). Two additional parallel troughs will be formed, dividing the disk into three segments. This will provide access to the top side of the liner for thermal cutting of the liner and removal of insulation in the kerf area. After the insulation is removed, a final thermal cut will be made to_ sever the remaining insulation cover plate, thereby penetrating through to the PCRV cavity. Prior to freeing the segments, they will be adequately supported. The l concrete / liner / insulation disk, after possible further segmentation, l will be removed and placed in a waste processing area for further sectioning, segregation and preparation for disposal. Removal of the top head section provides an access opening to the PCRV cavity. The final task of this activity is to set a PCRV-work platform on
- i. the ledge of the top head opening above the reactor core. This platform will be a rotating platform with openings to provide access to all sections of the PCRV. It is currently planned to have multiple workstations on this platform. This platform will also be provided _with underwater lights capable of being positioned to assist the workers with removal activities.
l 2.3-14 i l
6/28/91 .R0 POSED DECOMMISS10NING PLAN SECTION 2 During dismantlement operations, workers on the work platform will be protected from direct radiation and airborno contamination during removal of core components from the open PCRV. Radiation protection features include:
- Coro dismantlement will be performed underwatera shielding work"s and minimizing __ airborne particulate radioactivity._
PCRV Water Cleanup _and Clarification a system will_ st. _ p soluble radionuclides from the shield water. Tritium
,lnventory _ control is discussed in Section 3.3.2._2 of this plan.
The_ ventilation system will ensure a positive _ downward flow of air over the workers. Exhaust ducts under the _ work platform will carry air through a HEPA filter, then to_the existing plant ventilation system. Procedures and equipment for core dismantlement and _ operation of the work pl at form, will be prnvided to minimize radiation exposure to workers. All work will__ _ be_ performed in accordance with approv_ed Radiation Work Permits. 2.3.3.8 Dismantlina PCRV Core Como.pnents Following the removal of the PCRV top head and installation of the work ' platform, dismantling of PCRV core components will take place. These activities will include the removal of the reactor internals within the core barrel down to the CSF. Several reactor components are highly radioactive and will require special handling. These highly radioactive components (500 mrem /hr or greater) will be specifically identified and special handlinc procedures developed for their removal. PCRV core components will be disassembled and removed from the top down using manually operated grappling tools or power-operated hoisting equipment operated from the work platform (See figure 2.3 11). The initial task is to remove the 24 upper reflector keys, which must be accomplished in order to remove the side reflector blocks. The keys will be detached by removing he five nuts per key or by thermally cutting the keys. Removal of the approximate 5000 core componen_tt s (defueling elements, reflector blocks and core support blocks) will occur from 2.3-15
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 the top down. Removing the graphite blocks from the reactor will require several handling tools. The handling tools wil; be comprised of remotely engaged end effectors connected to a hoist by a cable. The cable and hoist arrangement will allow working with a shielding bell when required. A long handled pole, capable of being remotely attached or detached, will position the end effector on thc piece to be lifted. For (smantlement of the core components, l procedures will be written for the use of shielding, engineering controls, and respiratory protection equipment. The core components will be removed from the PCRV as described below.
- 1. Defueling elements: An end effector with a remotely operated grapple, similar to the fort St. Vrain FHM grapple mechanism, will engage the central lifting hole in the elements. As_ each element is lif ted from the core, water, will be removed from blind holes, and the elements dried.
The defueling elements are not activated and were uncontaminated when installed in the core. However, cross contamination is ( Jected to have occurred during reactor flooding. Contamination control procedures will be written for handling defueling elements. After deving, the
, elements will be packaged and loaded into a shipping container. There are 1482 defueling elements.
- 2. Replaceable and permanent hex re flec. tor blocks without Haste 11oy cans: An end effector with a remotely operated grapple will engage the lif ting hole. As each element is -
lifted from the core, w7ter will be removed from the blind holes, and the element dried. Six of these blocks are j metal clad reflector blocks and will require use of a
- shielding __ bell when they are lifted out of the PCRV.
Procedures will be written for handling these blocks. Af ter drying, the block will be packaged and loaded into i shielded shipping containers. There are 1693 replaceable and permanent hex reflector blocks without Haste 11oy cans.
- 3. Large side reflector blocks: A tool will lift the blocks by use of the two existing handling holes. The end effector will use remotely expandable collets to engage the holes. As each block is lifted from the core in a shielded i
container, water will be removed from the blind holes, and
- l. the block dried. Procedures will be written for handling l these blocks. After drying, the block will be sectioneda packaged, and loaded into shielded shipping containers.
There are 312 large side reflector blocks. 2.3-16
_ _ _ _ _ _ _.______.m. _ _ _ _ _ . _ _ . _ , _ 6/28/91 PROPOSED DECOMMISSIMING PLAN SECTION 2
- 4. Boronated spacer elements: A grappli 2 g tool will lift _the block to a pi_n_ dump station. There, the black will be inverted underwater to dump the boronated stainless steel pins into a collection cask. The block will be raised above the water to__ drain the blind holes, and then dried. l Procedures will be written for handling these blocks and i pins. After drying, the _ elemeats_ will be _ packaged and loaded into shielded shipping containers. The pins will be lifted in their cask, packaged; _ and loaded into shiel.ded shipping _ containers. There_ are_ 1152 boronated spacer elements.
- 5. flastelloy can hex reflector blocks: An end_ effector, with a _
remotely operated grapple, will engage _ the lif ting hole. Each element will be lifted into a shielding bell. Procedures will be written for handling these blocks. The block will - be packaged and- loaded into shielded shipging . containers. There are 2/2 Hast_elloy can hex reflector
. bl oc ks .-
l
- 6. Core support block, posts, end lower post seats: Grappling j tools will be used to remove the graphite components. As -
e%n component is lifted from the core, water will be removed from the blind holes and the piece dried. , _ Procedures will be written _for handling the components.- After drying, the larger core support blocks will be sectioned.- The components will be packaged and loaded into -- shicided shipping containers. There .are 61. core support blocks and 183 core support posts / lower post seats. Since the graphite blocks have been immersed in. water, appropriate steps will be taken in handling and pad aging the blocks to wipe, . drain or dry - the blocks as necessary u assure compliance with requirements spej,1 fled by 49 CFR,10 CFR 71, and disposal facility site criteria. Estimated contact dost ' rates for the graphite blocks is provided in Table 2.3-1. .Calculaions- to determine the contact exposure rate for various PCRV components were carried out in a conservative
, manner to estimate _ shielding requirements. For the purpose of developing a conservative estimate, the graphite block were .
considered to be-infinite in size and contain a uniform density of Co-60. The Co-60 values - are based on conservatively estimated cobalt impurity levels in the graphite. Exposure rates from other t core components like boron pins and Haste 11oy cans are based -on similar formulas, either infinite -slabs or point sources. Due to 2.3-17
. . _ . - - _2.__ . _ _ _ _ _ . .._.u.. __ - . - . . _
l PROPOSED DECOMMISSIONING Pl.AN 6/28/91 l SICTION 2 ! the much smaller size of the cans and_ pins, the contact exposure rates are very high. However, the exposure will decrease much more
,r.apidly with distance than for the larger graphite block.
After removing the graphite reflector blocks, the 24 lower core support block keys will be removed. The lower keys, which are made ; of Haste 11oy X, will have estimated radiation level 1 of 10 Rem /hr at l 1 meter. The lower core support block keys will be placed in a shielded container under water for movement to the radwaste area for packaging and disposal in a manner similar to that described above. The removal of the lower keys allows the core support blocks, posts, and post seats to be removed. This will be performed underwater l with grappling tools similar to those used in previous operations. 2.3.3.9 Removino the Core Barrel The core barrel and core barrel keys will be removed by thermal cutting. With the PCRV flooded above the core barrel and the water clarity established, mast mounted _ plasma equipment will be used to make vertical cuts around the core barrel. When the vertical cuts are_ complete, rigging will be attached to each piece before it is cut horizontally. The horizontal cut will then be made and the piece removed. As the core barrel is removed section by section downward, the keys will be cut loose from the core barrel and the PCRV. This method of rer. oval will continue down the entire lengt]h of the core barrel until approximately two feet of core barrel remains above the silica 1, locks. However, if radiological surveys in the core barrel indicate that the water level can be lowered ahead of the cuts, the core barrel and outer keys will be thermally cut above the water line. For either method identified above, fumes given off will be collected by installation of appropriate containment, exhaust fans, filters, and ducting in the work area. The core barrel will be cut into sections suitable for handling and packaging as activated radwaste. The core barrel will be sectioned and removed to a level just above the silica blocks on the CSF. ; 2.3.3.10 EREQYAl of the Core Sunfort floor 2.3.3.10.1 Removal of Silica Blocks. Cover Plates and insulation from the Core SupJort Floor _(_(}fl ! As currently planned, the water level will be lowered to just above the top surface of the CSF. Lowering of the water level in the PCRV 2.3-18 1 l-t
6/28/91 kJPOSED DECOMMISS10NING PLAN SECTION 2 will be accomplished by processing the water through the demineralizer system of the PCRV water cleanup and clarification system. The water will be analyzed prior to discharge. It is expected that the tritium level will be well below the unrestricted release limit for water discharge (0.003 microci/cc), since all of the graphite blocks will have been removed. The water can then be released directly from the site by directing it to the cooling tower blowdown water line, which is similar to the current liquid waste discharge route. As the water level is lowered in the PCRV, the walls will be washed down with clean water to remove residual loose contamination. A remotely-operated electro hydraulic ram hoe will be lowered into the PCRV to break up the silica blocks. A removable seal plate will be affixed to each of the 12 steam generator penetrations in the CSF to prevent loose debris from entering the steam generator modules. After the blocks have been fragmented, a bucket attachment will be affixed to the ram hoe to remove loose debris. The ram hoe controls and operator will be on a working platform above the CSF to minimize personnel exposure. The silica block debris will be removed in unshielded containers, since radiation levels are expected to be less than 500 mrem /l.r. The PCRV water level will be left sligMly above the CSF to minimize the potential for airborno releases during this operation. After the silica block debris has been remved, the insulation cover plates will be peeled up by the ram hoe with a sheet ripping attachment. The cover plates and any loose silica debris can be picked up, vacuumed, or scooped up at this time. The steam generator penetration seal plates will then be removed. 2.3.3.10.2 Removal of the Core Support Floor The CSF is a large disc approximately 29 feet in diameter by 5 feet thick and weighing 270 tons. The existing crane in the Reactor Buildint only has a capacity of 170 tons and cannot lift the CSF in one piece. Therefore, the CSF will have to be cut into pieces within the PCRV in order to be removed by the Reactor Building crane. This could require working in the radiation field that exists due to the CSF and the other components remaining in the PCRV, and with limited access on the sides and no direct access to the bottom of the CSF. It is therefore desirable to raise the CSF to the PCRV top head region for greater access and to reduce the radiation effects of other PCRV components. This will be accomplished by using a " strand jack _ing" system u which uses multiple cables hooked to the CSF and. attJched to hydraulic jacks _ posit _ioned 2.3-19 i
PROPOSED DECOMMISSIONING PL'3N 6/28/91 SECTION 2 on beams above the PCRV. The jacks will raise the CSF and place it i on supports on the ledge in the cavity. where the PCRV top head was cu,t and removed. Shieldino and radiological containment will be installed as necessary, and the CSF will be cut into segments small enough for handling by the Reactor Building crane. This method will provide greater access for the segmenting operation, and the work can be accomplished in a lower radiation field. A semi-remote in-air thermal cutting method will be used to cut the steam generator ducts from the CSF. An in-air thermal method will also be used to cut the CSF columns from the CSF. Stress analysis of the CSF columns will be utilized to determine the number of columns reluircJ to support the CSF at this stage of dismantlement. A lifting _: tructure will hold the CSF in place pt ior to cutting the remaining columns. Aftar cutting of u emaining CSF columns, the water level in the PC.' will be ra e '.o 4 level above the CSF. The CSF will then be l lifted into- tP rG cp head region and supports will be lowered into place. W LSE u 11 then be positioned on the supports while
;nside the PCRV tag P A area. The outside of the steel liner plate.
will be - scored wi th remainder of the floor will be sectioned while supported i: e PCRV top head cavity. -Temporary shield _ing will be used as necessary for all in-air operations. Fumes given-off will be collected by providing appropriate containment, exhaust fans, filter, and ducting to the existing plant ventilation. system. 2.3.3.11 Disassemblina the PCRV Low 6r Plenum During prior operations, the helium ducts connecting the CSF floor. to the 12. steam generators were severed and the CSF was removed from the PCRV and the water level lowered to below the CSF. Removal of the CSF will make lower plenum components accessible, including the steam generator primary assemblies, the helium diffusers, the CSF support columns, the lower floor, the lower plenum insalation and other miscellaneous components. The helium diffuser and shutoff valve assemblies will be removed
- . using techniques similar to those described in the following paragraphs for remcval of the steam generator primary assemblies.
The helium diffuser and shutoff valve assemblies will be l
- disconnected by remotely cutting the clamp at the connection of the l diffuser to - the lower floor. The assemblies will be rigged to the j Reactor Building crane, removed and transferred to the waste j- 2.3-20
~
l l 6/28/91 PROPOSED DECOMMISSIONING PLAN l SECTION 2 l handling- area for processing and disposal. All of the remaining l components in the lower plenum will be removed and transferred to ' l- the waste handling area for processing and disposal. Using' standard rigging techniques and devices in conjunction with the Reactor Building crane, the steam generator primary assemblies will be rigged to secure them before the final severance cut. Once the steam generator primary assembly has been rigged to the Reactor Building crane, it will be remotely disconnected by cutting the clamp at the connection of the steam generator shroud to the lower floor. Any - remaining instrumentation or connections between the steam . generators and the lower plenum will be severed remotely. When these activities have been completed, the steam generator may be removed from the PCRV cavity by the Reactor-Building crane. Because of the anticipated high contact dose rate associated with the steam generator primary assembly (economizer, evaporator, and superheater sections), a special shielded shipping container (Figure , 2.3-12) will be required. The following methodology will be employed to remove and ship the steam generator primary assembly. As the primary assemblies are lifted from the PCRV by the Reactor Building crane, the outer shroud and tube outer surfaces will be washed down to remove as much contamination and cutting debris as L possible, and will be allowed to drain as necessary over the PCRV L cavity. The steam generator primary assemblies will then be moved l to the truck bay. A shipping container will'be located in the truck bay to accept each steam generator primary module as it is removed from the PCRV. The shipping container will consist of a metal culvert section seven foot in diameter by 27 feet long,- The culvert section will be cut - in - half lengthwise -to _ provide a hollow hal f-cylinder. Structural supports will _ be welded to _ the half u section of culvert to provide structural support. L L . Support saddles will be mounted inside the culvert and serve a dual l purpose.- First, the saddles provide a means of attaching the steam generator primary assembly to the culvert and transmitting the load to the structural supports on the outside of the cuivert. .Second, the saddles will kaep the steam generator primary assembly centered in the culvert- with an annular space of about 8 inches between the inside diameter the culvert and the outside diameter of the steam generator primary assembly. The primary assembly will be lowered through the refueling deck access hatch to the~ truck bay, then transferred to the packaging and shipping area. The partial shipping container with the steam 2.3-21 i i I
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 generator will be moved to the packaging and shipping area, and the top half of the container will be installed. If required, the annular portion of the steam generator between the shroud and the i tube bundle support column may be filled with grout which will i encapsulate the tube bundle of the steam generator. In audition, grout may be pumped into the feednter and steam tubes of the l primary assembly. If necessary due to the high contamination levels, the 8 inch annular region between the outside of the steam generator shroud and the inside of the culvert will be filled with grout for shielding. The combined . weight of the shipping container, steam generator i assembly, and grout will be approximately 195,000 pounds. If actual l contamination levels in the steam generator primary assemblies are i lower than expected, the shielding grout in the annular space between the steam generator shroud and the containar may be omitted with a weight savings of about 56,000 pounds. With the steam generators removed, all of the significant radiation sources will have been removed. This will allow the PCRV vessel to be totally drained. The work remaining in the lower plenum includes the removal of the CSF support columns, the lower floor and the insulation and insulation cover plates on the PCRV liner and penetrations. These features will be removed utilizing hands-on tools ' and will- be processed for disposal. The Kaowool insulation removed in this activity- will most likely require removal of the absorbed water to assure compliance with shipping and disposal regulations. The removal of the absorbed water will initially be accomplished by pressing or squeezing the wet Kaowool, or other suitable drying techniques as required. During these final dismantling activiti.., the dose rates inside the PCRV lower plenum will be significantly lower than during previous operations since the largest radiation source, the steam generators, will- have been removed. It is estimated that the general area radiation level will be low enough to allow activities to be j performed in the lower plenum manually, which will increase l productivity and still be ALARA acceptable. 2.3.3.12 Final Dismantlina. Decontamination, and Cleanun Activitiet i The following activities are included in this task:
- 1. Scoring and cutting the PCRV sidewall insulation and liner.
- 2. -Cutting and removing the activated concrete in the beltline region of the PCRV (See Figures 2.3-13 and 2.3-14).
2.3-22
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2
- 3. Removal and/or decontamination of all remaining contaminated concrete.
, 4. Decontaminating the PCRV lower plenum liner.
- 5. Performing the final survey of the PCRV. i
- 6. Demobilization and decontamination of the PCRV D/D tools !
and equipment. l
- 7. Disposal of the water cleanup and clarification system.
l The activated concrete will be removed in sectional units from the side walls of the PCRV, with the attached liner and both layers of thermal "sulation intact as part of each unit. Diamond wire cutting has been selected as the method to remove the activated concrete sections. The thermal insulation, steel cover plates and steel seal sheets will be cut, and the liner plate will be scored by thermal methods -) before the concrete is cut with the diamond wire technique. This i prevents the diamond wire from entangling in the steel seal sheets ! and insulation. Tendons which must be removed for access of the diamond wire will be detensioned and removed. Other tendons detensioned to relieve compressive stress on the kerf of the diamond wire cut will be left in place. This was discussed in Section 2.3.3.5. I Circumferential tendons at the elevations of the horizontal cuts will be removed to provide a path for the diamond wire. The diamond wire cuts will be made in two steps from opposite directions, making a complete cut underneath the activated belt line concrete as shown in Figure 2.3-13. The inner ring of vertical PCRV tendon tubes, located 32 inches from the PCRV sidewall, are suitably positioned for removing the beltline i activated concrete (see Figure 2.3-14). Every third tendt, tube will be utilized for the radial and backside diamond wire cuts, thereby resulting in fourteen wall . sections to be removed. L Communications down the vertical tendon tubes, through the I horizontal cut, and up through the PCRV interior will allow threading of the diamond wire for the radial cuts to be made. Sections of concrete, liner and insulation that are approximately 3 feet thick, 8 feet wide, and 40 feet long will be produced and l rigged to the Reactor Building crane before the final back cut is l made between every third tendon tube. These sections will be moved to a radwaste processing area for further cuttir.g and preparation for disposal. 2.3-23 i
I i PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 I This cutting technique will remove a maximum depth of 32 inches at the tendon tube and a minimum depth of approximately 27 inches midway between the tendon tubes This minimum removal depth is adequate to meet the activation analysis requirements (plus uncertainties) described in Section 3.1.4.1. The water cleanup and clarification system will be dismantled and decommissioned-similar to balance of plan piping system. The system will be drained and the water processed as liquid waste as discussed in Section 3.3.2.2. The piping and components will be decontaminated, dismantled and packaged for disposal. The demineralizers will be the last items taken out of service. The demineralizer resins will be processed, packaged, and disposed of as radioactive waste. The demineralizers will be leased equipment, and will be decontaminated and pr.ckaged as necessary for return to the owner. Following the removal of the activated beltline concrete, a final cleanup and decontamination of the entire PCRV cavity will be performed. Decontamination methods may include conventional wiping techniques, scabbling, scari fying, vacuum sand blast, or a hydrolaser method, depending on the degree to which the contamination is fixed on the surface. A survey of the PCRV will be conducted to verify that free release criteria has been met. As disa.antlement activities proceed, guardrails, covers, barricades, caps, etc., will be placed as appropriate consistent with industrial safety considerations. Upon completion of PCRV activities, a top head closure along ' ith other appropriate penetration caps and guardrails. ' will be installed in compliance with good industrial safety practices. 2.3.4 Contaminated Balance of Plant System Dismantlement and Dscontamination 2.3.4.1 Introduction The decontamination and dismantlement of contaminated or potentially contaminated balance of plant systems will be done by either (1) decontamination in place, (2) removal and decontamination, or (3) removal and disposal as radioactive waste. Systems which are contaminated or potentially contaminated above releasable limits requiring decontamination or dismantlement include the following:
- 1. System 13 - Fuel Handling Equipment
- 2. System 14 - Fuel Storage Facility 2.3-24
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 l
- 3. Sy' stem 16 - Auxiliary Equipment
. _4 . System 21 - Helium Circulator Auxiliary Equipment
- 5. System 23 - Helium Purification System
- 6. System 24 - Helium Storage System
- 7. System 46 - Reactor Plant Cooling Water System ,
- 8. System 47 - Pur*fication Cooling Water System
- 9. System 61_- Decontamination System
- 10. System 62 - Radioactive Liquid Waste System i
- 11. System 63 - Radioactive Gas Waste System l
- 12. System 72 - Reactor Building Drain System
- 13. System 73 - Reactor Building Ventilation System
- 14. System 93 - Instrumentation and Controls Contaminated balance of plant decommissioning is scheduled to j l coincide with fluctuations in critical path PCRV activities to level 1 project- manpower and to minimize competition for use of plant J equipment.
I In general, contaminated or- potentially contaminated piping, ! components, structures, walls and ductwork will be dealt with in the , following_ manner. Potentially contaminated items will be surveyed l to determine acceptability for unrestricted free release or to i determine the cleanup required for release. Verification that plant systems or structures may be released for unrestricted use will be provided by a comprehensive radiological assessment that provides statistically signif' cant confidence levels for all plant systems. Since the plant systems cannot be altered for these detailed radiological surveys until the systems are no longer needed to meet NRC license requirements, the detailed surveys will be conducted during the implementation phase of the decommissioning project. The
.results of these radiological assessments will be used to determine the workscope required for final removal of contaminated or potentially contaminated systems and components.
The piping and equipment removal experience gained at the Shippingport Station Decommissioning Project demonstrated that contaminated or potentially contaminated piping and components can be quickly removed .by plasma-arc torch without compromising contamination controls when aided by portable HEPA filtered l ventilation units. Because of the relatively small volume of contaminated piping at Fort St. Vrain, however, the cost and support requirements of plasm -arc torch operations (setup, torch maintenance, and HEPA-filter changeout) may dictate the use of other methods, such as portable band saws, hydraulic shears, and alternate thermal cutting processes such as oxy-acetylene. Piping will be cut into segments of approximately equal length. As piping is removed, 2.3-25
6/28/91 PROPOSED DECOMMIMIONING PLAN
$fLT10N 2 as necessary to maintain a contamination envelope. The body of the FHM will be decontaminated and will be left on the refueling deck if release for unrestricted use limits are achieved. If further disassembly is required for release, the lead shot will be removed !
and the body will be segmented to segregate the contaminated i material from the uncontaminated components. The contaminated scrap j will be disposed of as described in Section 3.3 of this plan, i The reactor isolation valve exteriors will be surveyed and decontaminated by manual means. The valves will be removed from the op,erating floor and the lead shot removed. The shot is not expected to be contaminated or activated. The valve bodies will be disposed off-site according to Section 3.3 of this plan. The refueling sleeves will be surveyed and decontaminated by manutd means, then surveyed for release for unrestricted use. If they cannot be decontaminated, they will be disposed of as described in Section 3.3. The purge vacuum system will be removed and disposed of as described in Section 3.3. 2.3.4.3 System 14 - Fuel Storace Facility The fuel storage facility (See Fijure 2.213) consists of nine fuel storage wells (FSWs) constructed of carbon steel liners suspended in concrete pits. The FSWs were used for storini now and irradiated fuel during normal plant operation and may be used to temporarily store MCRBs or graphite reflector blocks during decommissioning. All fuel will have been removed from the Reactor Building prior to initiation of decommissioning activities. The actual contamination levels in the FSWs will be determined after the fuel has been permanently removed. When the FSWs are no longer needed, each of the nine inner storage wells will be decontaminated to the criteria for release for unrestricted use, surveyed, and the top access plugs replaced. The outer wells and the reactor - plant water cooling system are not contaminated and no outer well decontamination or dismantling is expected to- be required. The water cooling system piping at the bottom of each-pit will be cut open for survey. l Decontamination of the FSWs will be accomplished using a HEPA Following vacuuming, the wells will ~ filtered vacuum. be mechanically blasted with sand grit or cleaned using a hydrolaser. l Spent sand will be collected in catchments placed at the bottom of the well. The well drain pipe will provide water drainage if l 2.3-27 l
PROPOSED DECOMMISSIONING PLAN 6/28/91-SECTION 2 l hydrolaser operation is usod. After sandblasting or hydrolasing, the five standoff plates at the bottom of the wells will be removed minually. This will provide access for the final release surveys. Minor components will be shipped as radioactive waste rather than decontaminated. The well plugs will be decontaminated and replaced and sealed after the release surveys have been completed. ; 2.3.4.4 System 16 - Auxiliary Eauipment The auxiliary equipment consists of the Auxiliary Transfer Cask (ATC), (Figure 2.2.12), ten Equipment Storage Wells (ESWs), (Figure j 2.2-14), the Hot Service Facility (HSF), (Figure 2.2-15), and two l shielding adapters, (Figure 2.2-16). l The ATC was used to transfer the control rod drive assemblies 2 ; refueling sleeves and the shield plugs to and from the ESWs. The, ten ESWs are carbon steel structures embedded in concrete. They
,_were used to store the control rod drives and the refueling sleeves.
The Hot Service Facility is constructed of concrete and steel shielding and was used -for inspection, repair, maintenance, testing and decontamination work. Figure 2.2-16 is a general layout of the location of the various fuel handling and storage system components and associated auxiliary equipment on the refueling floor. All the components -of the ATC above the top base (32 ft. 11 in. above the operating floor) will be removed using the Reactor Building crane. A containment sleeve will be used to seal the contaminated ports in the cask and the hoist assembly floor as they are separated. The hoist cover and lift extension will then be lowered to- the operating floor and disassembled within -a contamination control envelope. The components will either be packaged and shipped for burial or to a licensed facility for processing and final disposition, or decontaminated and released for unrestricted use. The .emaining structure of the ATC will be decontaminated on site. The internal bore will be decontaminated using mechanical means such as sand blasting or hydrolasing to the criteria for release for unrestricted use. After internal decontamination, the crane will be used to lay the cask body over onto the floor for disassembly and decontamination of the bottom flange. When all surfaces meet the criteria for release for unrestricted use, it will be lifted by the crane and returned to storage on the operating floor. 2.3-28
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 The three shielding adapters will be decontaminated manually to the criteria for release for unrestricted use. The ten Equipment Storage Wells are internally contaminated and will be decontaminated to the criteria for release for unrestricted use and abandoned in place. Contamination levels in the ESWs will be determined when they are no longer needed. After the plugs are removed, the ESWs will be vacuumed using a HEPA vacuum assembly similar to that for the FSWs. After vacuuming the ESWs will be further cleaned using mechanical methods as necessary to reduce the contamination to the criteria for release for unrestricted use. After decontamination, the wells will be surveyed for release for unrestricted use. The top access plugs will be decontaminated, replaced a.d sealed. i Following final use of the HSF for decommissioning activities, all equipment (such as the manipulators and service platform sling) will j be removed. This equipment may either be decontaminated on site or j packaged and shipped to a licensed facility for processing and final j dispesition. The walls, floor, ceiling and remaining structural l components will then be decontaminated by sandblasting or ! hydrolazing. HEPA-filtered ventilation will be used to maintain a negative pressure in the HSF during decontaminations. 2.3.4.5 System 21 - Helium Circulator Auxiliaries The auxiliary equipment for System 21 was used to provide a supply of high-pressure water for the helium circulator bearing lubrication and a supply of purified buffer helium to prevent in-leakage of bearing water into the primary coolant helium. The major equipment items include buffer helium recircalators, heat exchangers, filters, pumps, helium dryers, chemical injection components, containment tanks, and compressors (See Figure 2.2-17). following the defueling of-the reactor, the helium circulator system l will no longer be used. It has no function in the decommissioning of the facility. The helium circulator auxiliary equipment is not expected to be , , contaminated above the criteria for release for unrestricted use l based on historical survey data. However, this system will be surveyed to determine the acceptability for release for unrestricted use following final use of the system. At present, no radioactive waste is expected to be genera ed from this system. 2.3-29
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 2.3.4.6 System 23 - Helium Purification System The helium purification auxilia_ry equipment consists of two trains and was used to assist in purification of the helium used as the primary reactor coolant. Most of the contaminated major equipment items are located in the PCRV top head and include filters, adsorbers, heat exchangers, dryers and piping (See Figure 2.2-18). System 23 equipment is located in the top head in eight wells. This equipment will be surveyed and a determination made whether to remove the equipment with the ATC or by manual means. All System 23 equipment located in the PCRV top head will be disposed of as radioactive wastes. After the wells have been emptied, they will be surveyed and decontaminated as necessary. The remainder of the helium purification system will be surveyed and l decontaminated or removed as necessarn 2.3.4.7 System 24 - Helium Storaae System i The primary purpose of the helium storage system was to provide for ; both storage and transfer of helium from the reactor vessel to the storage tanks. In addition, the helium storage system was used in testing the control rod reserve shutdown system and for various FHM purging operations. The primary equipment items include a helium
-transfer compressor, storage tanks, oil absorber, and high-pressure helium supply tanks (See Figure 2.2-19).
Following ' the defueling of the reactor, the helium storage system
- will no longer be used. It has no function in the decommissioning of the facility.
The 108 helium storage bottles and associated system components are not expected to be contaminated based on histe . cal survey data. The system will be surveyed following final use af the system. The results of this survey will be used to confirt the non-contaminated l- condition-or determine specific decontaminat.on steps for specific components. [ l . . l No radioactive waste is expected from this system. I 2.3.4.8 System 46 - Reactor Plant Coolina Water System The reactor plant cooling water system (see Figure 2.2-20) provides cooling water for process heat removal from all auxiliary equipment in the reactor plant. Three loops are provided that form the PCRV l 2.3-30 l
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 circuit- (liner cooling tubes), the PCRV auxiliary circuit (closed loop for various systems / components) and the service water circuit (open loop for various systems / components). The major equipment items include surge tanks, pumps, domineralizers, filters, heat exchangers, chemical injection (tank and pump) and recondenser chiller. The portions of the system external to the PCRV are not expected to be contaminated above the criteria for release for unrestricted use based on historical survey data. The system, however, will be cut-- open and surveyed following its final use. Clecnup requirements will be determined based on that survey. No radioactive waste is expected to be generated i t, the decommissioning of the portions of this system external to the PCRV. The reactor plant cooling water system loop supplying the PCRV will not be used for cooling of plant components during decommissioning. It will-be disconnected and isolated from the PCRV and from the FSWs before decommissioning of those systems occurs. Fifty percent of the PCRV cooling tubes will be cut and surveyed. The system is not expected to be contaminated above release limits. 2.3.4.9 System 47 - Purification Coolina Water System The purification cooling water system (two loops) provides cooling water to the helium purification system heat exchangers. The major components are pumps, expansion tanks, exchangers and associated piping (See Figure 2.2-21). This cooling' water system is not expected to be contaminated above the criteria for release for unrestricted use based on historical survey data. The system, however, will be surveyed following its final use. Cleanup requirements will be determined based on that survey and performed before final survey for release. There is not expected to be any radioactive waste generated in the decommissioning of this system. The purific3 tion cooling water system will be i solated from the helium purification system before it is decommissioned. The purification cooling water system has no other use during the decommissioning. l 2.3-31 l
. .. _ _ _ _ _ _ . . _ . . _ _ . . _ . _ .. _ _ _ ~. . _ _ .. .__ ___ _ _ _.
PROPOSED. DECOMMISSIONING' PLAN 6/28/91 SECTION 2
~
- 2.3.4.10 System 61-- Decontamination System The decontamination system consists of a water heater,=a drying air
~
heater, a filter, pumps, a solution tank and a chemical injection
- system (Figure 2.2-22)~ .
The decontamination system will be surveyed to determine the extent and- location of- radioactive contamination following final system use. The decontamination system components are small and- will- be removed -and packaged in LSA shipping containers with 'other contaminated components - and piping. The decontamination solution
-tank may be removed in one piece for shipment, or may be segmented and packaged in LSA shipping containers.
2.3.4.11 System 62 - Radioactive liauid Waste System The major equipment items in the Radioactive -Liquid Waste System include a waste sump (1000 - gallon tank), pumps, filters, two 3000 gallon receiver tanks, two demineralizers, and a 3000 gallon waste monitor tank (Figure 2.2-23). The liquid waste system is expected to be used for its original function during decommissioning operations- . Therefore, it will be one of the last systems to be decommissioned. A characterization' survey of the ~ radioactive liquid waste system
- will be performed, when the system is no longer needed, to determine the extent and location of radioactive contamination.
l The contaminated radioactive liquid -waste -system _ components are _ small- and include: ~ two liquid transfer pumps, two liquid waste sump. pumps,- twc liquid waste filters, and two liquid waste-demineralizers. The liquid waste monitor tank and the two liquid waste receivers may--be decontaminated -and -abandoned in pl ace, . . shipped as one piece containers,- or -segmented - and packaged- in LSA shipping containers depending on the extent and lccation of radioactive contamination. The-liquid waste sump will be considered for either (1) decontamination to the release for unrestricted use levels and abandonment, or-(2)-segmented and packaged as LSA waste.- 2.3.4.12 System 63 --Radioactive Gas Waste System p The major equipment items in this system include pre-filters, l- - filters,= exhaust blowers, - tanks (vacuum, surge, and drain), and L- compressor (Figure 2.?-24). 2.3-32
*,e .~ m , , - , , ,. =, w -w rm-e- - --
* *--'m -'t *'-~-
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 Following final use of the system, the radioactive gas waste system will be surveyed to determine the extent and location of radioactive contamination. The large components such as the two gas waste surge tanks and the gas waste vacuum tank may be decontaminated and abandoned in place, shipped as one-piece units, or segmented for
-gackaging and shipping. The other components are small enough to be shipped in LSA shipping containers with other contaminated piping.
Decontamination of these systems will be by- manT1 mechanical methods dependina on the levels of contamination f .und during the characterization survey. The system will not be used in the decommissioning of the plant. 2.3.4.13 System 72 - Reactor Building Drain System The major = equipment items include drain tanks, sump, pumps, piping and filters (Figure 2.2-25). Two gravity flow drains are provided to direct drainage from the Reactor Building equipment, piping and floor drains to either the radioactive liquid waste sump for potentially contaminated liquids or the Reactor Building sump for all other liquids. The drain system will continue to be used for its original function during much of the decommissioning work and will be one of the last systems to be decommissioned. When no loager required to remain operational, the system will be surveyed, and a decontamination and decommissioning decision will i then be made. Contaminated piping or- components will be either removed and shipped in _ LSA containers, or decontaminated to the criteria for release for unrestricted use and left in place. The portion of this-system that drains to the Reactor Building sump , is not expected to be contaminated. The portion of the system that drains to the radioactive liquid waste sump is expected to be contaminated and is included in dismantlement and removal plans. 2.3.4.14 System 73 - Reactor Buildino Ventilation System l l l The Reactor Building HVAC system services varicus areas of the Reactor Building with heated or cooled air. ' All ventilation air, whether outdoor or- recirculated, is filtered before distribution. In addition, the reactor plant HVAC system maintains building l differential pressure control. As shown in Figure 2.2-26 of the Proposed Decommissioning Plan, this system consists of several air handling units and filters. The only part of the system considered ! to contain - possible contamination is the React 3r Building exhaust l filters, hot service facility vent, and the analytical room vent. 2.3-33
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2-The reactor plant exhaust filters are composed of banks of moisture separators, HEPA filters and charcoal adsorbers. Based on historical data, the ventilation system is not expected to be contaminated above the release for unrestricted use limits. This system will t>e maintained during decommissioning to provide ventilation for decommissioning operations. The ventilation system . will be included in the characterization surven ' Following -final filter changeout and interior cleaning, ultimate deco.amissioning activities will be based on survey data taken at I that time. 2.3.4.15 System 93 - Instrumentation and Controls l The instruments and tubing to be removed or decontaminated originate at PCRV penetrations. These include thermometer penetrations, process and moisture instruments, helium circulator instruments, and helium vent piping. 1 Moisture monitors will be removed during dismantling the PCRV. All other instrument interfaces to contaminated or potentially I contaminated systems will be addressed when the respective system is l decommissioned. Those System 93 components will be either removed or verified to be below the limits for release for unrestricted use. All systems are scheduled for inclusion in the characterization survey. Contaminated system components will be decontaminated or disposed of as LSA waste. 2.3.5 Decommissioning Schedule l L The individual tasks making up the decommissioning effort have been delineated using a work breakdown structure (WBS) approach. Figure 2.3-15 is a schedule of the major decommissioning tasks which includes PCRV and balance of plant system dismantling and decontamination, and site decommissioning. This schedule is used as the top-level view of the project milestones and detailed schedules. Throughout the project, dismantling the PCRV is the critical path activity, with the B0P dismantling activities scheduled to coincide i with periods of reduced PCRV efforts ae a means of workload leveling. During the planning phase, work will be directed toward characterizing the site, preparing the decommissioning plan, and planning and writing the procedures and specifications for the implementation phase. The major activities and programs to be developed during the planning phase include: 2.3-34 l l
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2
- 1. Init2.1 sut haracterization
- 2. Dec M [U C _] planning
- 3. Work,gttcift stions and procedures
- 4. .__ quality _ assurance plan
- 5. Radiation protection program 6, Waste management plan 7 Proiect performance and control ine schedule depicts the planning phase occurring over an 18 month period t_and the actual dismantlement and decontamination activity at the site occurring over a 39 mnnth period.
2.3.6 Occupational Exposure Estimate The estimated cumulative radiation exposure for each major activity where the potential for worker exposure exists is provided in Table 2.3-2. The total cumulative occupational radiation exposure for the entire decommissioning pnjec_t _ in es_timated_t_o_be _433.. person-Rem2 due almost entirely to PCRV di smantiernent and associated waste handling activities. Actual measurements of radiation levels at individual work sites will be performed prior to commencing each individual task. These measurements may necessitate changes .a work procedures and projected exposures upon evaluation. The water shielding / clarity sysiem for the PCRV dismantlement activities will be the major exposure reduction device for the project. The 433 person-rem total exposure estimate will be used for planning purposes only and is not considered to be a restricting upper limit. Actual exposures will be controlled in accordance with ALARA principles (see Section 3.2). If projections indicate that the 433 person-rem estimate may be exceeded during the project, written notification will be provided to the Decommissioning Safety Review Committee (See Section 2.4.9) for assessment. 2.3.6.1 PCRV Dismantlement Cumulative Exposure Basis Estimates of the cumulative exposure for PCRV dismantlement were based on calculated activities for each activated component, estimated plateout activities for contaminated components, and the e_stimated person-hours required to complete the tasks outlined in Section 2.3.3 The cumulative exposure estimate for PCRV dismantlement also incorporates the benefit of utilizing the shielding introduced by the water system. 2.3-35
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 i The person-hours for completing PCRV dismantlement tasks were determined from discussions with craft supervisory personnel and j were based on the number of workers involved, the location of workers, and the duration of tasks in proximity to the high ra.diation fields. In general, the radiation fields are expected to vary significantly at different work locations, but for the purpose of est!maMng exposures, an effective exposure rate was assumed and applied to the entire work station c;ew. 2.3.6.2 Assumntions for Exposure, Estimates The assumptions and methods for calculating the cumulative radiation exposures consisted of: (1) utilizing the calculated exposure rates ' for reactor internal components; (2) utilizing projected times for completing each task where the potential ror radiation exposuret exists; and (3) utilizing engineering exper ence gained from similar projects in operating nuclear plants. The issumptions used in tiie calculation were as follows:
- 1. For PCRV operations only, the time workers spend at the work station radiation environment was assumed-to be 50% of the time scheduled to complete the task.
- 2. The " crew averaged" radiation levels were determined by assuming the total exposures estimated for completing a task would be uniformly distributed among the crew.
- 3. The graphite reflector blocks will be removed without the use of shielded transfer casks. (Shadow shields, distance and long-handled tools will be used while removing highly radioactive components.)
- 4. The activated boronated pins will be loaded into shielded containers under water - and transferred to the HSF for processing.
- 5. The PCRV water level will be maintained such that the general area exposure rate on the work platform will be
- less than 2.0 mrem /hr.
l
- 6. The workers will be trained in ALARA principles and good work practices to minimize occupational exposures.
i 2.3-36 I
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 2.3.6.3 BOP Cumulative Exposure Basis i s 4 The estimate of cumulative exposure for the balance of plant (B0P) systems _ external to the PCRV was based on historical survey data and estimates of person-hours required to perform the tasks outlined in Section 2.3.4 of this plan. In _ general, radiation levels for BOP systems measure less than 2.0 mrem /hr. lherefore, the removal of contaminated BOP components and decontamination of remaining j structures is not expected to contribute significantly to worker i exposures. 1 I l l l t 2.3-37
c- -.%-.,a .,.4 .- Am4 Ja -+m a-wE a . 4 - - - - - , . (4.-d- F 4-L J-- A w4atup-4.- .AL--aa4.-A .-W.___._2 PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION INTENTIONALLY LEFT BLANK 1 l l. l l l 2.3-38
k 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 TABLE 2.3-1 ESTIMATED CONTACT DOSE RATES FOR CRAPHITE BLOCKS ESTIMATED CONTACT TRANSPORT NO. OF DOSE CASK GRAPHITE BLOCK DESCRIPTION I!LQCXS Rale RE0VIRED
- 1) Defueling Bincks 1,482 <1 mR/hr No
- 2) Hexagonal Reflector 1,687 500 mR/hr No Blocks with no Hastelloy fini
- 3) Large Permanent 312 <30 R/hr Yes Reflectors
- 4) Reflector Keys 24 <100 mR/hr No
- 5) Side Spacer Blocks (a) with Boron Rods 1,152 30 R/hr Yes (b) Boron Rods removed 1,152 <3 mR/hr Yes (c) Boron Rods 276,096 6D R/hr Vei
- 6) Bottom Reflector Blocks (a) with Hastelloy Cans 272 300 R/hr Yes M
l (b) Hastelloy Cans 500 mR/hr Yes removed l (c) Hastelloy Cans 19,452 10,000 R/hr Yes l 7) Core Support Block 24 1,000 R/hr Yes Hastelloy Keys l' 8) Core Support Blocks (61) 244 15 mR/hr No i and Posts (183)
- 9) MCRBs (Non-control rod) 6 300 R/hr Yes
PROPOSED' DECOMMISSIONING PLAN 6/28/91-SECTION 2 TABLE 2.3-2 PROJECTED PERSON-REM EXPOSURE FOR THE FORT ST. VRAIN DECOMISSIONING PROJECT ESTIMATED WORK ACTIVITY
- PERSON ** PERSON HOURS EXPOSURE 2.3 PCRV DISMANTLEMENT AND DECONTAMINATION Initial- preparation / disassembly '23,733 7I4 Remove PCRV concrete top head 20,578 2_02 Dismantle _PCRV-core and core barrel 49,368 157.3 Remove core support floor, barrel & insulation 9,213 103.4
-D/D PCRV Lower plenum 16,103 59.9 Final PCRV _ dismantlement, decontamination and cleanup <
15,047 17.7
'SUB-TOTAL 134,042 366~
2.4 CONTAMINATED B0P D/D & WASTE PACKAGING Ini ti al - preparat i on/charac teri zation 7, 2_7_9 0.25 Dismantle /decon operations 58,684 1.4
- - SUB-TOTAL 65,963 1.65-2.6 WASTE PREPARATION, PACKAGING, SHIPPING -
AND DISPOSAL 33,055 65.4 TOTAL '233,060 433 l
- Person-hours only for= those tasks where the; potential for-measuring radiaticn exposures exists
** Ex30sure time (worker efficiency) is estimated to be 50% of '
scleduled work time for PCRV tasks where the potential for radiation exposure exists. I
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- g a . g gf . DU = o . g.i. . ._ . $ E 7 s 5 -A r Figure 2.3-15 Schedule of Major Decomissioning Tasks L PROP 0 SED DECOMMISSIONING PLAN 6/28/91 : SECTION ? p INTENT 10HALLY LEFT BLANK I i l 1 I . . . . ~ , - ...,.__,_..___._.,,-_ _ _,_. . ,, ...,___,.,:.,.,.._..,_,. l 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 2.4 DECDMMISS10NING ORGANIZAT10N AND RESPONSIBILITIES 2,4.1 PSC Commitment Public Service Company of Colorado (PSC) is fully committed to compliance with the existing license and applicable regulatory requirements during all phases of the fort St. Vrain decommissioning. PSC's commitment to the safe decommissioning of the facility will be accomplished with diligence and quality service. Corporate principles, policios, and goais will be followed to ensure performance excellence, management competence, and high standards in every facet of the decommissioning project. 2.4.2 PSC Dec missionena Oraanization and funC110D1 The PSC Decommissioning staff for the Fort St. Vrafn Nuclear Generating Station and the -interface with the Westinghouse team is shown in figure 2.4-1. lhe manpower level is approximately 60 people including the key staff members and all performance icvel people. Overall onsite control and responsibility for all decommissioning activities for both PSC and contractor personnel rests with the PSC Program Manager for Decome.issioning. Within the PSC organization, four main groups rep >>rt to the Program Manager for Decommissioning. The groups consitt of tne- Project Assurance . Manager, facility Support Managst, Operations Manager, and Engineering Manager Decommissioning. Contractor reporting requirements and lines of authority are identified in Section 2.5. The PSC Program Manager for Decommissioning interfaces directly with the Westinghouse Project Director for decommissioning activities. During the decommissioning process, PSC will retain responsibility for the 10 CFR 50 license and therefore will maintain the following responsibilities: - 1. M rall management oversight of all decommissioning project activities.
- t. Sole point of contact with all. regulatory agencies within the. Federal, State and local governments.
- 3. Overall responsibility for all licensing activities.
- 4. Overall management of those plans and programs required to comply with licensing requirements, including: access control, radiation protection. Decommissioning Emergency '
Response Plan, . fire protection, Quality Assurance,. maintenance and operation -of existing plant systems, training and configuration management. 2.4 1 PROPOSED DECOMMISSIONING PLAN 6/28/01 SECTION 2 The key decommissioning staff members perform the functions described in the following subsections. 2.4.3 l'SC Corporate Vice Pres 14taL_thslear_Qperation] The Corporate Vice President, Nuclear Operations, provides the corporate support at the corporate executive level and has the authority and responsibility to ensure that all activities to carry cut decommissioning are performed safely and within applicable regulations. The Vice President Nuclear Operations shall hee a a nimum of fifteen years executive experience in waste management, decontamination and decommissioning, and nuclear operations. He must have a formal education in an engineering or physical science field. Knowledge in the areas of regulation and compliance, decommissioning techniques, and applied rt:iiation protection programs are required. in addition, a background of knowledge with respect to NRC and DOE is desirable. 2.4.4 Program Miu1Ager for Decemissionina The Program Manager for Decommissioning is directly responsible to the Vice President, Nuclear Operations. The Program Manager for Decommissioning coordinates and oversees all decommissioning activities. This position provides direction to the support groups to ensure radiological anc industrial safety, compliance with regulatory requirements, cost effectiveness, and interfaces for PSC Labor Relations of the decommissioning project. The Westinghouse Team Project Director will report to this position. The Program Manager for Decommissioning shall have a minimum of ten years responsible plant experience with formal education in an Engineering or Physical Science field. A significant technical background to have good working knowledge of plant principles of operation, maintenance and engineering principles. Additional knowledge in the areas of regulation and compliance, decommissioning techniques and applied radiation protection programs are required. 2.4.5 Pro.iect Assurance Manaaer The Project Assurance Manager is responsible for Quality Assurance oversight (including QA reviews, audits and monitoring 1 surveillance) activities), licensing and regulatory compliance, and overall industrial safety. To ensure the independence of the QA function, this position reports to the Vice President, Nuclear 2.4-2 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 . Opvations,onqualityassuranceandlicensingmatterslas_ indicated byp dotted line_ in Figure 2.41). The Project Assurance Manager gports directly to the Program __ Manager for Decommissioning for administrative direction and implementation of_the Quality Assurance Plan, coordination and direction for licensing activities, and coordination direction of the industrial safety _ progam (as indicated by_the solid line in Figure 2.4-1). The Project Assurance Manager shall have a minimum of _ five years experience in a responsible position that includes coordination, direction and supervision of- personnel, a formal education in an Engineering or Physical Science field, and a working knowledge and understanding of plant design and operation and construction practices is required. A balance of experience in quality assurance related activities and in regulatory / compliance requirements are preferred. 2.4.6 Facility Support Manaaer The facility Support Manager is responsible for the Radiation Protection, ALARA, Access Control and Training programs. This , position is also responsible for managing support areas of records control and retention, PSC training, PSC materials and facilities, The Facility Support Manager has the overall responsibility for the Radiation Protection Program described in Section 3.2 of this plan, and shall serve as the PSC Radiation Protection Manager. The facility Support Manager represents the formal line of communication and authority between Fort St. Vrain and the Westinghouse team for radiation protection matters related to decommissioning. This individual will be responsible for ensuring that the Radiation Protection Program and procedures meet the goals and standards estah11shed by PSC management and the governing regulatory agencies. The facility Support Manager will also be directly responsible for the Radiological Environmental Monitoring Program and the Decommissioning Emergency Response Pl an (Secti_on 9). This individual will meet the _ qualification contained in NRC Regulatory Guide 1.8 " Qualification and Training of Personnel for Nuclear Power Plants" (Ref. 6). The duties and responsibilities of the Facility Support Manager with respect to the Radiation Protection Program are described in further detail in_S_ection 3.2.3 of this plan. The facility Support Manager shall have a minimum of five years experience in a responsible -position that includes coordination, l direction and supervision of personnel. A formal education in Engineering or the Physical Sciences or the equivalent experience in 2.4-3 l PROPOSED DECOMMISS10filNG PLAN 6/28/91 SECTION 2 a science or engineering subject is preferred. Formal training in radiation protection is required. 2.4.7 Operations Manaaer The Operations Manager is responsible for the overall conduct and management of operations and maintenance functions. These responsibilities include system operations, testing and surveillances, system maintenance, lay-up and turnover. The Opz'ations Manager shall have a minimum of eight years of responsible power plant experience of which five must be nuclear power plant experience, including coordination, direction and supervision of personnel. A thorough working knowledge and understanding of pl ant design and operation and maintenance functions (including instrumentation and control maintenance activities) are required. 2.4.8 Enaineerina Manaaer. Decommissioning The Engineering Manager, Decommissioning, is responsible the administrative and technical functions of the decommissioning project. Responsible areas include management of contract work, technical assistance, evaluation and administration of contract changes, project scheduling and overall general contracts. This position is also responsible for the general oversight of field work, preparation of engineering evaluations and coordination with operations. The Engineering Manager, Decommissioning, shall have a minimum of a Bachelor's Degree in Engineering or the Physical Sciences and have a minimum of five years of professional level experience in nuclear services, nuclear plant design and operation, including coordination, direction and supervision of personnel. A working knowledge and understanding of decommissioning techniques, scheduling and contract W inistration is required. 2.4.9 Decommissionina Safety Review Committee The DSRC is composed of the Program Manager for Decommissioning (Chairman), facility Support Manager, Engineering Manager Decommissioning, Operations Manager, Project Assurance Manager and the Westinghouse Team Project Director. This committee reports to the Vice President of Nu:iear Operations. The function of this committee is to monitor decemmissioning operations to ensure that they are being performed safely. This committee will review and i 2.4-4 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 audit major decommissioning operations dealing with radioactive material and radiological controls. In addition, they will review work gecification_s and administrative procedures, reportable occurrences under 10 CfR 20 and 10 CfR 50, and changes made in , accordance with 10 CfR 50.59. i 2.4-5 1 PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 INTENTIONALLY LEFT BLANK 2,4-6 6/28/91 PROPOSED DECOMMISSIONING PLAN SEC110N 2 PRf51D(NT & C(0 PUBl!C SlRvlCE COMPANY OF COLORADO VLCC PRESIDENT NUCLEAR OPERAil0NS . DECOMM15510NING SAF(1Y RIVl[V COMM11 Tit PROGRAM MANAG[R VESTINGHOUS[ DCCOMMISSIONING PROJfti DIRfCTOR PROJECT ASSURANCE OPERATIONS ENGINEERING FACILITY SUPPORT MANA0(R MANAGER MANAGfR MANAGtR DECOMMI5510NING RADIATION PROTICTION MANAGER l l l l l l figure 2.4-1 PSC Decommissioning Organization PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 l l INTENTIONALLY LEFT BLANK 1 i f i I l- 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 2.5 CONTRACTOR RESPONSIBillTIES 2.5.1 ht+.inahouse Team Organiza11pA and Functions This section describes the responsibilities, work scope, and qualifications of the Westinghouse team that will perform the decommissioning. Since PSC retains overall licensing responsibility, the PSC a id Westinghouse team interf ace will be structured to clearly demonstrate PSC compliance and control as required. Westinghouse will, as a minimum, obtain approval of top level project procedures, plans and programs that could have an impact on license compliance or safety. 2.5.2 hstinahouse Team Scope of Work Based on a competitive bid selection process, PSC selected the Westinghouse team to perform the decommissioning, dismantlement and decontamination of Fort St. Vrain. This team is an affiliation of Westinghouse and MK-ferguson Corporations. Westinghouse is the overall lead organization for the program. Within the Westinghouse team, Westinghouse will be responsible for overall project management. in addition, Westinghouse will also be responsible for decommissioning engineering, licensing support, and will provide the project quality plan throughout the project. The overall responsibilities of each of the Westinghouse team members are summarized in Figure 2.5-1. The Westinghouse Quality Assurance organization, responsibilities and reporting lines of communication are described in Section 7 of this plan and depicted in Figure 2.5-2. MK-Ferguson will provide the site labor, labor management and supporting infrastructure for decommissionir3 The fort St. Vrain decommissioning project will be conducted in two phases (See Section 1.2.5). A breakdown of the Westinghouse team work scope is provided in Appendix 1 of this plan. 2,5.3 Oraanization of the Westinahouse Team This section identifies the Westinghouse team organization and their responsibilities in the fort St. Vrain decommissioning project. The Westinghouse team combines many years of successful experience in the design, construction, operation, and decommissioning of commercial and government-owned nuclear facilities. Westinghouse and MK-Ferguson have a strong commitment to the project and to operating in a safe, environmentally sound, cost-effective and responsible manner. 2.5-1 PROPOSED DECOMMISSIONING PLM 6/28/91 SECTION 2 The Westinghouse Team Organization Chart, figure 2.5 2, shows the interrelationship of the positions within the Westinghouse team project organization. Descriptions of the responsibilities and qualifications are provided in the following paragraphs. 2.5.3.1 fort St. Vrain Westinahouse Pro.iect Directar The Project Director is the most senior and responsible management position in the Westinghouse team project organization. The Westinghouse team Project Director will provide a single point of contact for PSC on the decommissioning effort. The Project Director reports to the PSC Program Manager for Decommissioning and is fully responsible for Westinghouse team personnel, plant safety, prevention of environmental occurrences, quality assurance, project integrity, costs, schedule, efficiency, and technical output of the overall program. The Project Director will be responsible for project implementation for both project phases and will have full authority to administer Westinghouse team resources. The Project Director reports directly to the Vice President of the Westinghouse Energy Systems Busines , Unit. Other duties and responsibilities include the following:
- 1. Establish project manning requirements, organizational structure, scope, and necessary levels of expertise.
- 2. Select and manage the project staff.
- 3. Ensure that the project schedule and budget are properly detailed and defined.
- 4. Direct the set up of all project control programs, operating plans, and technical services. -
- 5. Ensure that the project meets applicable regulatory standards.
- 6. Direct all phases of site work, including preplanning, l mobilization, training, temporary facility erection, decontamination and dismantling activities, conversion, and project closcout.
- 7. Ensure that all work activities are carried out according to the project standards of safety, quality, and reliability.
- 8. Direct team members, technical services, and operations and control activities _for entire project.
- 9. Enforce adherence to the project policies and procedures.
-2.5-2 1 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 2.5.3.2 Technical Services Manasitt The Technical Services Manager is responsible for engineering, licensing support, and environmental control activities. The Technical Services Manager reports directly to the Westinghouse Project Director. A requirement for a Technical Services Manager is based on the need for quality engineering and technical support services to ensure successful on site operation. The Technical Services Manager is responsible for the technical services organization, which will perform a wide va"tety of work in support of the overall facility decommissioning. Initial efforts include completion of the decommissioning plan, tooling design, and procedure and process definition. Following completion of this preliminary work, emphasis will shift to pru iding technical support in design of tools, development of pr>cedures, operation, waste processing and waste management, radiochemistry and health physics for the D/D effort and waste disposal activities. Duties of the Technical Services Managar will consist of managing department manpower and funding allocations. The Technical Services Manager also ensures that technical aspects of the project are done in a safe, disciplined, and quality manner. Other duties and responsibilities of th; Technical Services Manager include the following:
- 1. Perform the engineering necessary to support work package development, including tooling design and material lists.
Ensure proper review / approval of engineering documents, purchase orders, field design changes, and other documents, as required.
- 3. Maintain engineering team to provide design engineering
, support. L 4. Incorporation of technical requirements, methods, j regulations and procedures for waste processing activities into engineering and operational documents.
- 5. Provide. input to establish an engineering schedule / budget and continuously monitor cost / accomplishments against established budget and schedule.
6, Advise the project manager of changes in plans or changed conditions affecting costs or schedules. 2.5-3 PROPOSED DECOMMIS$10NING PLAN 6/28/91 SEC110N 2 2.5.3.3 Qn g ations ManagqC The Operations Manager is responsible for the performance of all dismantling, decontamination, and conversion activities in a safe, i disciplined, and quality manner. 1he Operations Manager reports directly to the Westinghouse Project Director. The Operations Manager is responsible for the safe, disciplined management of the decontamination and dismantling activities, waste processing, and facilities management by implementing the radiation protection program, and quality assurance plan throughout his department. The Operations Manager will also ensure that adequate ' short-term internal planning is accomplished and that this planning is in agreement with the project master schedule, as well as PSC goals and objectives. The Operations Manager is also charged with accurately identifying the personnel and physical resources required to complete all production tasks, and for identifying and integrating the operations department scope of work, budget, and schedule. Other duties and responsibilities of the Operations Manager include the following:
- 1. Manage the day-to-day activities of the project team at the site as well as the subcontractors. These activities include decommissioning, decontamination, waste handling and site maintenance.
- 2. Prepare work packages and specs for the modification or removal of plant structures.
- 3. Prepare and manage work packages for decontamination, dismantllnt and asbestos = removal and disposal.
- 4. Estimate fie ) activity costs and record them,
- 5. Prepare procedures and specs for modifications, deenergizing, or removal of electrical power, lighting and switchgear, and systems operation.
- 6. Prepare and manage work packages for the modification and removal of piping, components and systems, including IIVAC.
- 7. Supervise project procurement.
- 8. Manage project site industrial rafety program _and medical facilities.
- 9. Ensure that training is-provided to the work force.
- 10. Provide work package document control.
- 11. Acquire and manage craft labor for decommissioning, decontamination,-waste handling, and site maintenance.
- 12. Manage and interface -routine project activities with PSC, including as a minimum clearance and system turnover.
2.5-4 I l I ) 6/28/91 PROPOSED DECOMMISSIONING PLAN t SEC110N 2 l 2.5.3.4 Eroject Radiation Protection Manaaer The _ Project Radiation Protection Manager has the responsibility for the implementation of the fort St. Vrain Decommissioning Radiation Protection Program. This includes Radiation Protection Program development under the direction of the PSC Radiation Protection Manager, implementotion and essuring compliance with applicable regulations. The Profect Radiation Protection Manager will serve as the co Chairman of the AL ARA committee with the PSC Radiation Proter. tion Manager. The Project Radiation Protection Manager reports directly to the Westinghouse Project Director. The Project Radiation Protection Manager will be respons_ible for the approval of the content of _the radiation protection training p_rograms, for the selection and approval _of all radiationJrotection staff members, and for the___ review and__ approval of all radiation protection procedures. The Project Radiation Protection Managel will be qualified in _ accordance _ wl_th NRC Regulatory Guide _ l.8 _" Qualification and Training of_ Personnel at Nuclear Power Plants" ' (Ref 6). The duties and responsibilities of the Project Radiation Protection Manager are discussed in further detail in Section 3.2 and include the following:
- 1. Implement and maintain an effective radiation protection program as required for the Fort St. Vrain decommissioning.
_ 2. Establish and maintain an effective ALARA program.
- 3. Establish and maintain a program that minimizes the volume of radioactive waste and that ensures safe transportation and disposal of radioective waste _ materia 1.
- 4. Provide radiation protection input to decommissioning planning.
2.5.3.5 Project Control Manaaer The Project Control Manager is responsible for all scheduling, project _ control and reporting systems, and _the integration of reports at the -project level. The Project Control Manager reports directly to the Westinghouse Project Director. The Project Control i.. nager ensures that all relevant information
- regarding co:,t/ schedule integration and control is available in the l
pruper forrnat (:ither summary or detailed data). The Project Control Manager will also address any schedule variances, including outliniq the problem, providing potential solutions, assessing 2.5-5 PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 input and reporting analysis results. This will be accomplished by the set up of an appropriate level budget / schedule control system l broken down into definable work packages. Necessary subsystems, such as collection, reporting, and analysis, will combine to form a total system that will provide a significant management tool. Other duties and responsibilities of the Project Control Manager include the following: ]
- 1. Prepara and evaluate productivity data. l
- 2. Supervise overall cost and scheduling functions.
- 3. Prepare management reports.
- 4. Prepare and coordinate detailed activity schedules.
5, Forecast cost and analyzes trends.
- 6. Evaluate schedule impacts and formulate alternate plans as necessary.
- 7. Ensure time and labor studies are done to determine costs on specific operations.
- 8. Maintain interface with the PSC administrative and scheduling functional groups.
- 9. Establish and maintain a records retention system. I
- 10. Establish and maintain a management information system. l
- 11. Establish and perform audit activities, as required (financial, schedules, progress).
- 12. Prepare and verify project invoices.
2.5.3.5 Quality Assutaatc_tianaggt The vestinghouse team Quality Assurance Manager reports to _the Vice President and General Manager, Energy Systems Business Unit. .on gualitLassurance matters and to the Westinghouse Project Director for administrative direction and implementation of the Quality Assurance Plan. The Quality Assurance Manager is responsible for implementation of the Quality Assurance Plan described in detail in Section 7 of this plan. 2.5.3.7 Coroorate Commitment Executive corporate management.of each Westinghouse team member will continue to monitor fort St Vrain decommissioning project progress through direct lines of communication and reporting (See figure 2.5 2). The Westinghouse Project Director will report to the Vice President and General Manager, Energy Systems Business V91t. The Fort St. Vrain decommissioning project has been assigned a high
- priority within MK-Ferguson. The Operations Manager is responsible .
2.5-6 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 for MK ferguson's project scope, and reports to the Executive Vice President, MK forguen. 2.5.4 )!as11n9hMLle_hELQ1talifica11011s an1Lxffriente lhe Westinghouse team has extensive and comprehensive experience performing all the activities necessary to successfully decommission fort St. Vrain. Expertise in each of the disciplines essential to successful completion of the fort St. Vrain project is evidenced in the matrix of projects and experience provided in Table 2.5 1. Westinghouse and MK ferguson have shared joint work experience on the following projects:
- 1. Shippingport Decommissioning Project
- 2. WEPC0 (Point Beach) Steam Generator Replacement Project
- 3. WPPSS Design and Construction Projects
- 4. AEP (D. C. Cook) Steam Generator Repair Project
- 5. AlW/A4W Nuclear Reactor Prototype
- 6. Shippingport Plant Modification
- 7. Savannah River This shared experience will promote a cohesive approach, productivity, close integration under Westinghouse project management, and consistency in the fort St. Vrain project.
2.5.4.1 Site Release Experients The Westinghouse team has successfully achieved the unrestricted release of the facilities listed in Table 2.5-2. MK Ferguson currently is the DOE contr.ctor for the Uranium Mill Tailings Remedial Actions Project (UMTRAP). This effort, which began in 1983, encompasses 22 sites and approximately 700 vicinity properties in ten states. The scope of this project includes verifying the effectiveness of remedial actions. These remedial actions are documented in the vicinity property completion reports. Westinghouse personnel were contributors to the document " final Consolidated Implementation Plan for Site Release, Rev. 1". This plan was used to direct site release activities at the Shippingport Decommissioning Project. Westinghouse personnel have actively participated in final site characterization and firal report preparation activities. 2.5-7 l l l PROPOSED DECOMMISS10NillG PLAN 6/28/91 SECTION 2 2.5.4.2 Rdialngcal Protttlion Exonitnn The Westinghouse team has extensive experience in designing and implementing effective radiation protection programs for projects like fort St. Vrain. The team has developed or played a major role in developing radiation protection and ALARA programs for routine and outage activities. Examples of facilities where plans have been successfully implemented include: - Limerick Generating Station - Peach Bottom Atomic Power Station Nine Mlle Point Unit 1 - Waterford 3 Steam Electric Station - Salem Generating Station - Rancho Seco Nuclear Plant - Pilgrim Nuclear Power Station E.1. Hatch Nuclear Plant - Shoreham Nuclear Power Station Westinghouse and MK-ferguson are experienced in developing ALARA programs and evaluating methods of reducing occupational radiation exposure (ORE).
- 1. Westinghouse ALARA Programs:
- Developed dose models for tracking exposure by task, identi fying areas where improvement is needed, and developing cost effective solutions. - Sponsors the Radiation Exposure Management seminar, an international symposium for communicating and discussing ORE topics among plant health physicists and ALARA engineers from plants designed by Westinghouse and others. - Participates in Electric Power Research Institute (EPRI) cooperative programs, aimed at identifying and reducing radiation sources in nuclear plants. - Performed radiological assessments for various plant sites and licenses, including National Nuclear Corporation, Mitsubishi Heavy Industries and NIRA/SOPREN. - Provided onsite ALARA consultation and coordination during San Onofre and Millstone steam generator sleeving, Connecticut Yankee refueling, Krsko (Yugoslavia) steam 2.5-8 6/20/91 PROPOSED DiCOMMISS10NING PLAN SECTION 2 TABLE 2.5 1 RELEVANT WESTINGHOUSE TEAM EXPERIENCE '%g,k ;l3i;> l ;l. i j > i .j > ._. >;3i -. _4_ i 't . _ {_N h j N N . N' b { - - _ 7 . -- 4 5 diIN !. ]
- 50 5 _
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- il* * *i
%(,%s. mi > > x ;' > ; > i j a, vi ai e { b ,{ @ @.[ M ,{ M ,{ 1 A ,{ M ,{ { N 4'.. j %e' _h_#_-.. i F i 5 _F.4i S i .. -) i # yi 5i h g o. _h h h . 5 . '5.. ._. _mN . h h hlWh N 5h 5 h- > Wi,l5.__ _5.. lY$ 5 $l5 $ W _.N$ 5i $ ,N. ___ h h: >' h hl h' h > - . - - . . _ = . - - - - - _ .-._-..-+-. . . - . . , - . - %# _; .l >_ _ > . - \x _. x u_. 5 !x vu _ x ..\>__ _ kil_)i X jyi )i R $;F 'N i _._ , x. wil>; s )i 5 * ;>i ti xi x __._ ivi vil . i s a ir l t v2 w w -{ g m ? , B p $5 gl b$ BhIB su d 4i eib g eiosw n g: 2pt eu A s; t egg eh i a o n" 2;a v wy l PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 l l t b INTENTIONALLY LEFT BLANK ( l i i l e _ r [ s . _ . . . .,_m...i--. ,._....,-s .i..,4_,. . . . . - . . . . . , ...-.m...s,-.-... ..m~-.. , .- . , . . . , . , _ , . . . , ...,......w,,,m,,, m I 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2 *(5fth"# 0V$[ (h(RGY [x[CUTIVE SY11(M$ BU$lh($$ UN!f VlC[ FREllDthi VP AND G(h(RAL PAh4GIR K-f(RGU$0h PSC l PROGRAM MANAG(R ! OCCOMMISSION!NG l l l (~~' l FORT $1, VRA!N , C) I Pn0;tet olatcica l I l .. I I T[CHNICAL FROJECT CONTROL OPERATIONS SERVICE $ MANAGER MANAG[R MANAG{R l l l 1 QUALITY ASSURANCE MANAGER i Figure 2.5-2 Westinghouse Team Organization Chart ,*y,--.->-pwww- ,, p ---en,- yr -w,tg,r-e-+w----m-e-w m----mrw- ww-vr -e s PROPOSED DECOMMISSIONING PLAN SECTION 2 1 1 1 l INTENTIONALLY LEFT DLANK > 4 t t l .a-.-_.-..--~-.---.----.--.---. _ - . . - - - . . . - - - . - - - - - _ - 6/28/91 PROPOSED DECOMMISSIONING plt.N SECTION 2 2.6- 1 RAINING PROGRAM All personnel who may require access to the work areas or a radiologically controlled area 1RCAJ associated with the fort St. Vrain Decommissioning, whether PSC employees, contractor employees, or visitors, shall receive appropriate training commensurate with i the potential hazards to which they may be exposed. 2.6.1 General Emoloyee Training . General Employee Training (GET) will be provided to all personnel j assigned to decommissioning on a regular basis. This training will l include:
- 1. Project orientation / Access Control.
- 2. Introduction to Radiation Protection.
- 3. Fire protection, j
- 4. Quality assurance, i
- 5. Industrial safety, i
- 6. Emergency response.
2.6.2 Radiation worker Trainina Basic radiation worker training shall be provided to persons who routinely work in radiologically controlled areas of the project. Basic radiation worker training covers a large range of topics including:
- 1. Fundamentals of radiation.
- 2. 31ological effects of radiation.
- 3. External radiation exposure limits and control.
- 4. Internal radiation limits and controls.
- 5. Contamination limits and controls.
- 6. Management and control of radioactive waste, including waste minimization practices.
- 7. Response to emergencies.
- 8. Radiation Protection Program.
-In addition to a classroom presentation of the topics identified -above, participants in basic radiation worker training are required to participate in the following demonstrations:
- 1. The proper procedures for donning and removing a complete set of protective clothing (excluding respiratory l protection equipment).
l 2.6-1 I PROPOSED DECOMMISSIONING PLA' 6/28/91 SECTION 2
- 2. The ability to read and interpret self reading dosimeters.
- 3. Proper procedures for entering and exiting a contaminated area, including use of proper frisking techniques. ;
1
- 4. An understanding of the use of Radiation Work Permits 1 (RWPs) by working within the requirements of a given RWP. ,
Personnel completing basic radiation worker training are required to pass a written examination on the material presented. Completion of this training qualifies an individual for unescorted access to radiologically controlled areas of the project. 2.6.3 Soccific Job Training , Specific job training will be provided to selected workers based upon the:r job assignments and their need to know. Training programs shall assure the following:
- 1. Personnel responsible for performing activities are instructed as to the purpose, scope, and implementation of applicable controlling procedures. .
- 2. Personnel performing activities are trained as appropriate, ;
in principles and techniques of- the activity tving - performed.
- 3. The scope, objectives, and methods of implementing the training programs are documented.
Examples of this training are as follows: ,
- 1. Respirator training.
Personnel whose work assignments require them to use respiratory protection devices will receive training in the < devices that they will be required to use. The training program follows the requirements of 10_ - CFR_ 20.103 and Regulatory Guide 8.15. " Acceptable Programs for Respiratory Protection" (Ref. 7). Training -consists of a classroom session and a simulated work session. In addition, fit testing and medical evaluations are required in order to use respiratory protection devices.
- 2. Asbestos worker training.
- 3. Mock-up-training.
l 4. Training on use of special tools or equipment.
- 5. Work package briefings.
2.6-2 . . ~ . , _ - __ - - - . . . . _ _ _ _ _ _ . - _ _ . - - . _ _ . . . . _ . _ . - - . _ . - - - , - - - - - - - ..- - . - . . - . _ - - - . _ j 6/28/91 PROPOSED DEC0ffilSS10Niku PLAN SEC110N 2
- 6. Fire watch. !
- 7. Radioactive material transportation training. '
- 8. Health physics technician training.
Training and qualification of health physics technicians and supervisors will be conducted in accordance with ANS 3.1-1981 (Ref. 8), 2.6.4 Non Radiation Worker Indoctrination All non radiation workers who require access to a radiologically controlled area will receive an appropriate indoctrination prior to being permitted access into RCAs. This indoctrination will include l as a minimum:
- 1. The requirement that non radiation workers remain with I their escort at all times and follow the directions of the
-escort.
- 2. A description of the radiological conditions and required controls of the area to be entered.
- 3. The purpose and proper use of dosimeters, including how to read a self-reading dosimeter and the appropriate exposure <
limits.
- 4. Potential emergency cituations and proper actions to take in such events.
2.6.5 Radiation Protection Staff Trainina Training and qualification requirement of the radiation protection staff during tha decommissioning have been established through the guidance provided by NUREG 0761 (Draft) " Radiation Protection Plans for Nuclear Power Plants" (Ref. 9). Section 3.2.4 of this plan describes the radiation protection staff training and qualification requirements. 2.6.6 Trainina Records Records of training will be maintained which will include trainees name, date of training, type of training, test results, authorization for protective equipment use, and instructors name. A list of qualified instructors will be maintained. 2.6-3 L L - .. . PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 Training records will be organized in several ways, to allow either a listing of an individuals qualifications or listings of rersonnel due for retraining. The interval between initial training and retraining will be identified, as appropriate, in training procedures. t 2.6-4 l 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 2
2.7 REFERENCES
FOR SECTION 2
- 1. USAEC Regu'.atory Guide 1.86 " Termination of Operating Licenses for Nuclear Reactors", June 1974.
- 2. " Evaporation Atlas fcr the 48 Contiguous United States", NOAA Technical Report NWS-33, Department of Commerce,1982.
- 3. NRC Regulatory Guide 1.60 " Design Response Spectra for Seismic Design of Nuclear Power Plants", Rev.1, l_973.
- 4. NRC Regulatory Guide 1.61 " Damping Va_ lues for Seismic Design of Nuclear Power Plants , 1973.
- 5. NRC Regulatory Guide 1.143 Des ign Guidance for Radioactive Waste Management Systems, Structures, and Components Installed In Light Water-Cooled Nuclear Power Plants", Rev. 1, 0:tober .
1978. } 1
- 6. NRC Regulatory Guide 1.8 " Qualification and Training of l Personnel at Nuclear Power Plants", Rev, 2, April 1987.
- 7. NRC Regulatory Guide 8.15 " Acceptable Programs for Respiratory Protection," Octobe_r_1976.
- 8. " Selection, Qualification ane Training of Personncl for Nuclear Power Plants," ANS 3.1-1981.
- 9. NUREG-0761 "Radiat' m Protection Plans for Nuclear Power Reactor Licensees" (Oraft), March 1981.
2.7-1
M_- b h-A #e4.4 4. -y _ - - .M.-- +n W * -tMe +4es.h-.-- r +-h- -1 44~esa4 h. 4*ay- J_$A6Wa4 5 A* 4 +#aA 5x=_Ax a._%c4 A AA.4sa ___, I PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 2 1 i i l l 4 INTENTIONALLY LEFT BLANK h i l 2.7-2
. _. . .-_: _ __ ~ .. . _ _ .--. _ . - - -
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 SECTION 3 l PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY l 3.1 FACILITY RADIOLOGICAL STATUS 3.1.1 Facility Operatina History Construction of the fort St, Vrain Nuclear Generating Station was authorized by the NRC by issuance to Pubile Service Company of Colorado a provisional construction permit on September 17, 1968. Construction was complete in December 1973. Fuel was loaded and l l nuclear criticality achieved on January 31, 1974. After a prolonged ! period of startup testing, low-power operation and plant l modifications, the plant was committed for commercial operation July ! 1, 1979. Full power operation was achieved on November 16, 1081. In August 1989, the plant was shutdown due to control rod drive l problems . Due to these problems, as well as additional mechanical l and financial concerns, the PSC Board of Directors decided not to restart the plant. This announcement was made August 29, 1989. PSC
.has now commenced defueling and has begun preliminary plant closure
, -activities. 1 During the operational history of the plant there have been no spills or release of radioactive effluents resulting in significant residual radioactive contamination either onsite or offsite. However, there have been a few routine plant operations that may L have resulted in residual radioactive contamination in areas which are inaccessible. l Specifically, the fuel Storage Wells (FSWs) and Equipment Storage l Wells (ESWs) on the *efueling floor were used to store spent fuel and highly- radioactivs components. Over the years of transferring l various components and spent fuel, it is anticipated that high l levels (e.g. > 5,000,000 dpm/100 cm2) of loose surface contamination
- will have accumulated on horizontal surfaces. The lower portions.of L these wells are inaccessible at present (See Figure 2.2-13). At various times throughout plant history, the Hot Service Facility (HSF) has also had levels of loose surface contamination measuring greater than 5,000,000 dpm/100 cm2 Periodic decontamination of the HSF was typically performed using water, and as a result, crud traps' may have been created in inaccessible areas. To date, no crud traps have been identified in accessible areas containing drain piping from the HSF (See Figure 2.2-15).
3.1-1
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 1 3.1.2 Radiation Soynn 3.1.2.1 Description of Instrumentation and Survey Techniaues In August 1990, radiation and contamination surveys were performed in the Reactor _ and Turbine Buildings. These surveys focused on identifying of the major contributors to radiation levels above background and areas containing both fixed and loose surface contamination. The radiation surveys were performed using the instruments listed ; below and were performed to detect gamma radiation only. Radiation j levels were reported in units of mrem /hr. Radiation levels are ; measured either as general area (approximately waist high), contact (typically measured with the probe within one inch of the radiation source), or at a distance from the contact reading (approximately 18 inches from the source of radiation). Instrument Scale Ranae/ Switch Settinas Eberline R0-2 mrem /hr 0-5/0-50/0-500/0-5000 Ludlum 19 microR/hr 0-25/0-50/0-250/0-500/0-5000 Fixed contamination surveys were performed using Eberline RM-14/15
' friskers which measure beta-gamma contamination in counts per minute
.(cpm) using a scale of 0 - 500 cpm with switch settings of x1, x10, x100, and x1000 (x1000 available on the RM-15 only). Conversion from cpm to disintegrations per minute-(dpm) uses a conservative counting efficiency of ten percent. Survey results are reported in units of dpm/ probe area, which is approximately 15' square centimeters.
The counting of_ wipes. (or smears) for loose surface contamination was performed using either a Tennelec LB 5100 or a Harshaw TASC-12-A-6 which analyzed beta activity only. Although the minimum detectable activity (MDA) will vary slightly on a daily basis, the typical MDA for these instruments is approximately ten dpm. Results are reported in units of dpm/100 cm2. _ (100 cm2 is the area over-which the wipe / smear is taken,_ approximate?y four square inches). Radiological surveys in the past have shown that alpha contamination (both fixed and loose surface) is not present above natural background levels at Fort St. Vrain. Spot checks for alpha contamination are performed on a routine basis to confirm this. Generally, the results of these surveys demonstrated that greater than 95% of the plant areas have radiation levels corresponding to 3.1-2
1 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 natural background (in the 0.004 to 0.032 mrem /hr range). In the results summarized in Table 3.1-1 at the end of this section, only those areas with radiation levels above these background levels are noted. Additionally, fixed contamination levels are generally less than 1000 dpm/15 cm2 and loose surface contamination levels of less than 1000 dpm/100 cm2 Most survey results are less than 100 dpm/100 cm2 In some locations, tritium may be present as fixed contamination. Due to the low energy beta activity emitted by tritium (Eavg - 0.005 MeV), normal survey methods will not detect the tritium and therefore actual tritium levels are not measured. Fixed contamination is typically imbedded within the first few centimeters of concrete surfaces. The contribution to area radiation levels from facilities, rooms and structures where various components are undergoing routine plant maintenance activities were not included in the survey results due to the temporary, transient nature of such activities. Figures 3.1-1 to 3.1-19 provide specific results of these area radiation surveys. Table 3.1-1 provides a summary of the survey results with a description of the major contributors to the radiation levels. Reactor and Turbine Building elevati.ons are shown in Figure 3.1-20. Where results are not listed, contamination l- - and/or radiation levels are not greater than background levels. l Systems which are potentially contaminated are identified in Table 3.1-l~ by system number for each elevation on which they are located. 3.1.2.2 Turbine Buildino Survey Results l l General area radiation levels throughout the Turbine Building are primarily due to natural background. Contamination levels (both E fixed and loose) are less than 1000 dpm/100 cm2 in all locations and. generally less than 100 dpm/100 cm2 Piping from the potentially internally contaminated Systems 11 and 73 extends from level 7 (El. 4829') to the roof of the Reactor Building. 3.1.2.3 Radiation Sources outside the Reactor and Turbine Buildinos Radioactive material s are stored on a temporary basis inside Sea-Vans and cargo trailers. The locations of these trailers are indicated on Figure 3.1-21. Varying amounts of radioactive-meterials may be stored in these trailers, but external radiation levels are typically less than 0.2 mrem /hr. 3.1-3
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 The only permanently contaminated area outside the Reactor and Turbine Buildings is the Compactor Building directly east of the main cooling tower (see Figure 3.1-1 and 2.2-1). General area radiation levels vary from 0.2 to 0.5 mrem /hr primarily due to residual contamination inside a radioactive waste compactor. Loose surface contamination levels are generally less than 100 dpm/100 cm2 The compactor contains loose surface contamination of 50,000 dpm/100 cm2 and fixed contamination levels of 50,000 dpm/100 cm2, There are two concrete bunkers in the Compactor Building which have loose surface contamination levels of 5,000 dpm/100 cm2 and fixed contamination levels in the first few centimeters of the concrete averaging approximately 20,000 dpm/100 cm2. The presence of tritium is also suspected in the fixed contamination of the bunkers. This building is also used for staging cf radioactive wastes (including liquids) and materials. Piping associated with the Radioactive Liquid Waste System (System
- 62) also runs underground from the exit point from the Reactor Building to the main cooling towers. Sample results of oil collected in an associated oil separator have occasionally shown trace amounts of tritium, C0-60, Cs-137 and Cs-134.
Routine surveys do not indicate any radiation or contamination levels above background in the Radiochemistry Laboratory located in the Technical Support Building (See Figure 2.2-1), although small amounts of radioactivity may later be found in drain piping from this facility to the Radioactive liquid Waste System (System 62). 3.1.5 Current Environmental Radioloaical Status 3.1.3.1 Beta-Gamma Radiation in Surroundina Environs The environmental radiological status of the site and surrounding areas has been monitored during the entire pre-operational, operational, and post-operational phases of the plant through the Radiological Environmental Monitoring Program (REMP). This program includes surveillances in surrounding areas to gather environmental data in the following areas: external gamma activity levels, air sampling data, water sampling data, milk data, aquatic pathways, and food products. Sample locations are located near the site boundaries and in outlying areas. Details of the results of these surveillances can be found in Reference 1 and in past REMP reports, which are provided annually to the NRC. During the spring and summer of 1990, additional data were taken to further characterize the site. Soil samples were taken inside and 3.1-4
6/28/91 PROPOSED DECOMMISSIONING PLAN
- SECTION 3 outside the protected area, gamma radiation surveys were performed inside the protected area, and downwind air samples were taken with respect to the predominant wind direction (from the NE).
Environmental radiation surveillance data from all past REMP reports and the recent characterization data indicate that the predominant source terms found above natural background levels are due to Chernobyl and past nuclear weapons test fallout. External radiation sources to area residents are due to naturally-occurring background radiation and atmospheric fallout. The recent characterization data included the exposure rate from gamma-ray emitting radionuclides and were measured using thermoluminescent dosimeters (TLD). The TLD stations were constructed at 72 different locations inside and outside the controlled area boundary. Each station contained packets with two chips of CaF2 (Dy), which are _ identical to those used in the REMP. The measurement period for the TLDs was 92 days. The mean of the two chips in each station was used to determine the mean exposure rate._ The overall mean exposure rate from the TLD packages was 0.32 mrem / day. This value is not statistically different from the mean value found _in the 1989 REMP report (Ref. 1) of 0.38 mrem / day for the Fort St. Vrain facility area. Reference 1 indicated that since the inception of power production by the reactor, there has been no l detectable increase in the external exposure rate due to planned or i unplanned reactor releases. The concentrations of gross beta activity due to the combination of o naturally occurring radionuclides and fission product radionuclides l was determined from air _ samples at two locations downwind from the L predominant wind direction. A particulate filter for gross beta l analysis and an activated charcoal cartridge for 1-131 or noble gas radionuclide analyses were_ in the sample line. Tritium in atmospheric water vapor _ is collected passively by silica gel at each of these locations. Sampling methodology was identical to that utilized in the REMP. Fort St. Vrain operational- technical specifications no longer require measurement of gross alpha activity. -Gross beta activity measured in air particulates was l= principally due to naturally occurring radionuclides or from soil resuspension. The mean weekly activity air concentrations measured at the northern and southern monitors were 16 femtoCuries(E-15)/m3 These concentrations are comparable to those found in the REMP program. Past REMP data has shown that there has never been a significant difference observed between facility and reference sites (Ref. 1). It is concluded, therefore, that based on the current radiological data and past REMP data, the reactor air effluents of 3.1-5
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 particulate fission products or activation products are not a source of dose commitment for the Fort St. Vrain environs population. 3.1.3.2 Sqil, Samples Soil samplcs wer< v at 124 locations inside and outside the controlled area dere taken at each location from a depth of ten centirr iti area of 95 square centimeters. Two samples were take. - 'ch location to produce a sample size sufficient to fill a on ; art volume. Samples were dried, ground to a constant density, and sealed in the quart container. After a in.ee week period, ead container was counted using Ge(Li) gamma-ray spectroscopy to determee the activity concentration of important fission products, activotion products and naturally occurring radionuclides. Deep core samples (taken at approximately 12 percent of the soil sample locations) were taken to approximately 150 centimeters in depth. The core samples were collected in polyethylene tubes, which were frozen and sectioned off to obtain samples at various depths. The deep core samples were analyzed using the same techniques as the soil samples. Results of the soil samples indicate the precence of statistically significant Cs-137 concentrations. These concentrations are due tc, world-wide fallout remaining from the United States, USSR and Chinese nuclear weapons tests, and the Chernobyl accident. This is supported by the fact that the Cs-137 concentrations are the same in the entire front range of Colorado and other reactor-generated fission products or activation products were not present in the samples. 3.1.3.3 Results of REMP Surveillances Tritium is the only radionuclide that was detected in concentrations above background in any effluent pathways that could be attributed to reactor operation. Since tritium is released as tritiated water, the dilution by the surrounding hydrosphere is significant. Elevated level s of tritiated water (Ref. 1) were detected in downstream surface water samples on occasion, but the yearly nean values of downstream surface water was not statistically greater than upstream concentrations. Tritium concentrations measured in milk were all less than the lower limit of detection (LLD). ilowever, slight increases in the downstream tritium levels, which were discussed in the 1986 REMP report, showed that the radiation dose commitment that can be calculated as a result of the increases 3.1-6
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 was found to be negligible as compared to natural background radiation dose rates. The REMP program has, over the years, been shown to be of adequate scope and sensitivity to detect any accidental releases from Fort St. Vrain operation. It is concluded that the dose commitments calculated for the closest inhabitants or other parts of the nearby ecosystems due to reactor operations are negligible. In addition to the REMP data, the most recent characterization data both inside and outside the controlled area boundary supports this conclusion. The negligible release of radioactivity from Fort St. Vrain is due to its unique gas cooled design. 3.1.4 Radionuclide Inventory 3.1.4.1 Activated Components within the PCRV An activation analysis (Ref. 2) was performed for the PCRV and associated internal components and is provided in Appendix II. The analysis was performed to estimate the isotopic composition, magnitude and extent of residual radioactivity which could be present in the PCRV after the end of operations. The actual operating history of the plant was used in the analysis by considering the total effective full power days (EFPD) generated by the plant until its shutdown in August 1989. The analysis consisted of three sections: (1) neutron flux estimates in the PCRV; (2) activation analysis of the PCRV and internal components; and (3) calculation of gamma dose rates (in air) inside the PCRV due to non-removable (fixed) components. 3.1.4.1.1 Computer Codes The activation analysis required the use of several computer codes and various input data libraries. The ANISN m . (Ref. 3) was used to determine the neutron flux throughout the reactor core and outward through the reflectors, helium flow paths, insulation, PCRV liner and PCRV concrete. The activation of selected components within the PCRV was then determined using the REBATE computer code (Ref. 4). Finally, gamma doses (in air) within the PCRV were calculated using the REBATE, ANISN and other data manipulation codes. 3.1.4.1.2 Material Compositions Material compositions of components were determined from a variety of sources. In most cases, material compositions were identified 3.1-7
d'.0 POSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 from component drawings which referenced standard material specifications. Assumptions for the number densities of trace elements, such as europium (Eu), cobalt (Co), and niobium (Nb), were based on design _ manuals, previous analytical investigations and recent regulatory guidance (Ref. 5). _In general, the reactor internals are made.of carbon steel, graphite or concrete. Few major components are made from stainless steel or Inconel. Details of the actual compositions and trace element assumptions are found in Reference 2, which is provided in Appendix II, 3.1.4.1.3 Comoutational Models The analysis was computed using three one-dimensional models: (1) Radial - core center line outward through the PCRV side (see Figure 3.1-22), (2) Axial Up - core center line upward through the PCRV top head and (3) Axial Down - core center line downward through the core support floor. Below the CSF, activation was assumed to be insignificant with the exception of the activation of the top of the steam generator modules due.to neutron streaming effects (Ref 6). Descriptions of each of these models are provided in the following paragraphs. 3.1.4.1.4 Radial Model The " Radial" model consists of the following components:
- 1. Side removable reflectors
- 2. Large permanent side reflectors
- 3. Boronated side spacer blocks i- 4, Core barrel l S. Kaowool insulation and insulation cover plates
- 6. PCRV liner and cooling tubes-
- 7. PCRV concrete and rebar
- 8. Reflector keys and. carbon steel metal shell L The removable (hexagonal) graphite side reflector elements are L located just outside of the active core (See Figure 3.1-23). These L elements will be removed during defueling. A typical block is approximately 31 inches in height and 14 inches across the flats.
These reflectors were modeled in the neutron flux calculations, but not in the activation. calculations. 3utside the removable side reflectors are the large permanent side i reflector blocks. These blocks are irregular in shape (See '-igure 3.1-23) and have-an average volume of 3.358E+05 cc each. 3.1-8 .
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 Between the large permanent side reflectors and the core barrel are the boronated side spacer blocks (Figure 3.1-23). These graphite blocks contain boronated stainless steel pins which provided neutron shielding during power operations. The number of pins vary with respect to the blocks position relative to the active core region. The core barrel (Figure 3.1-23) is located just beyond the boronated spacer blocks and serves as the lateral restraint of the fuel end reflectors. The barrel is constructed of carbon steel, in th"ee sections which vary in thickness from 2.25 inches to 2.75 incies and is approximately 29 feet in height. The core barrel was modeled in two sections because the upper 12 feet is constructed of a slightly different material than the bottom two sections. Continuing outward from the core barrel, Class A Kaowool insulation and cover plates (Figure 3.1-24) cover the inside of the PCRV liner. The in.ulation is a ceramic fiber material and the cover plates are constructed of carbon steel. The PCRV liner (Figure 3.1-24) is a 3/4-inch carbs, steel plate vessel in the form of a right circular cylinder, 75 feet in height with an inside diameter of 31 feet. Carbon steel cooling tubes, welded to the outer (concrete) side of the liner, provided cooling to the concrete during power operations. The liner and cooling tubes were modeled homogeneously for the activation analysis. The PCRV serves as containment of the nuclear steam supply system (NSSS). The concrete walls vary in thickness from 9 feet to 15-1/2 feet, and the vessel is approximately 106 feet high. The PCRV was modeled as a homogeneous mixture of concrete and rebir. Two final components modeled were the carbon steel side reflector block keys (Figure 3.1-23) which connect the core barrel to the la-]e side reflectors and the carbon steel metal shell for the top most large side reflectors (half length reflectors). 3.1.4.1.5 Axial UD Model The " Axial Up" model included the following components:
- 1. Metal clad reflector blocks (MCRBs)
- 2. Region constraint devices (RCDs)
- 3. Lower orifice valve assembly
- 4. Kaowool insulation and insulation cover plates
- 5. PCRV liner and cooling tubes
- 6. PCRV top head concrete and rebar 3.1-9
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 The MCRBs are located on the top most level of the core area (See Figure 3.1-23) and- provided structural stability and neutron shielding during power operations. All blocks are hexagonal in shape, approximately 15 inches in height and -14 inches across the f lats. The 37 central column MCRBs (Figure 3.1-E5) are constructed ' primarily of stainless steel. The remaining 270 MCRBs, with and without coolant holes (Figure 3.1-26), are constructed of carbon steel. The RCDs- (Figure 3.1-27) provided restraint of fuel regions during power operations. The RCDs are located on top of the MCRBs, keying fuel columns between regions together. The triangular main body of the device is made of carbon -steel, approximately 5 inches thick. The " legs" of the RCD are approximately 7 inches in length and are composed of Inconel. The orifice valve assembly (Figure 3.1-28) is located just above the , central column MCRB. The lower portion of this assembly, primarily I composed of carbon steel, was modeled as part of the axial up mo&l. The final three components (Kaowool/ cover plates, PCRV liner / cooling tubas and PCRV concrete /rebar) were modeled as previously discussed in the radial model, i l 3.1.4.1.6 Axial Down Model The " Axial Down" model included six components:
- 1. Removable bottom reflectors
- 2. Core support blocks and core support posts
- 3. Silica block insulation
- 4. Kaowool insulation and insulation cover plates
- 5. CSF liner and cooling tubes L 6. CSF concrete and rebar l-Directly beneath the active core are removable - hexagonal bottom L reflectors (Figure 3.1-29). These include graphite reflectors, L graphite reflectors containing boronated Hastelloy-X cans and i
graphite transition' reflectors which channel coolar.t from the bottom-reflectors to the core support blocks. The core support blocks and core support posts (Figure 3.1-29) are permaneat components which i lie directly below the removable reflectors and act as support for l the fuel and reflectors. Three layers of Class C silica insulation (blocks) are located above the cover plate /Kaowool which are just above the CSF liner (See Figure 3.1-30). The CSF liner is a 3.1-10
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 3/4-inch carbon steel liner encasing the five foot thick concrete /rebar core support floor (See figure 3.1-29). 3.1.4.1.7 A_clivation Analysis Results The results of the activation analysis are summarized in Table 3.1-2. Detailed results can be found in Appendix II. The nuclides of importance as well as the total estimated radionuclide inventory for activated components inside the PCRV are listed in the table. The dominant nuclides for metallic components are Fe-55, Co-60, Ni-63 and Mn-54. Traces of Nb-94 and Fe-59 are also present in some metallic components. The dominant gamma emitter in the stainless steel components was determined to be Co-60, although Nb-94 is also present. Due to the high concentrations of Co-60 in the boronated spacers blocks, these components are the primary dose contributors inside the PCRV. The activity In graphite components is dominated ';y tritium and Fe-55, which were generated due to impurities in the graphite. Due to the large volume of graphite and the high curie content of tritium and Fe-55, these components are the largest contributors to the overall radionuclide inventory. No credit was taken for the migration of tritium out of the graphite. The Kaowool insulation and silica blocks were determined to have fairly low activities. The carbon steel cover plate contains almost all the activity in the Kaowool/ cover plate assemblies. The silica block activity is dominated by Fe-55. The PCRV concrete /rebar mixture contains many activation products due to the presence of trace elements. In the short term, Co-60 is the dominant gamma emitter, while Eu-152 and Eu-154 are the dominant long term gamma emitters. The nuclide contributing most to the total activity is Fe-55. Other nuclides present in lower activities were: Cs-134, Ca-45, Ag-110m, tritium, C-14, Fe-59, Ni-59, Ni-63, Nb-94, Mn-54, and Ca-41. Specific details of the calculated isotopic breakdown for each activated component can be found in Appendix C of Appendix II. As indicated in Table 3.1-2, the majority of the activity in the concrete is contained in the first 1.5 feet in all directions. Table 3.1-3 indicates the estimated required amount of concrete which must be removed to achieve the recommended release limit for unrestricted use (5 microR/hr above background). The activation 3.1-11
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 ; i analysis predicts that at five years after shutdown, api aimately l 24 inches of concrete will require removal from the PCRV sidewalls to achieve a dose rate of 3.4 microR/hr. The 24-inch depth is based ! on the assumption that the dose rate in the PCRV will be due to contributions frca, the top head, CSF and PCRV sidewalls. The top head and CSF will be removed during decommissioning, leaving only the PCRV sidewalls contributing to the dose rate within the PCRV. In this case, a dose rate of 5 microR/hr is predicted to be achieved l when approximately 23 inches of PCRV sidewall concrete is removed. j Dose rate estimates for r :h direction within the PCRV are listed in -l Table 3.1-3 and the total dose inside the center of the PCRV (in air) is the sum of the dose rate for all three directions. Table ! 3.13-also indicates the estimated dose rate contribution (in air) ! l for various stages of component removal for an individual located in l the center of the PCRV.
]
I The activation analysis modeled the PCRV concrete and rebar as a l homogeneous mixture. la the a_ctual rebar configuration, two sets of l rebar lay just outside the PCRV liner and the remainder of the rebar lays at least 1.5 feet beyond the inside ring of tendons. The two inner sets of rebar will be removed during decommissioning and the l outer sets of rebar will not be activated. Therefore, the assumption of a homogeneous mixture at a depth of 23 inches of concrete is conservative. l A sensitivity study of the contribution of the rebar to the overall l dose rate was performed to assess this conservatism. The rebar number densities were removed from the -homogeneous mixture and the activation analysis was rerun. The resulting predicted dose rates at five years' after shutdown, with 20 inches and 22 inches of concrete removed, are- 6.15 microR/hr and 3.14 microR/hr, respectively. Therefore, the actual predicted removal depth is approximately 21 inches rather than 23 inches to achieve a dose rate in the 5 microR/hr range. i 3.1.4.1.8 Activation Analysis Verification l l The thermal neutron flux and trace element abundances are considered to be the two most important potential sources of uncertainty in the activation analysis. An action plan has been developed (Ref. 7) to assess the accuracy and sensitivity of these parameters and the impact to the overall activation analysis results. This action plan, which is currently in progress, is designed to validate neutron flux and veri fy material composition assumptions. The action plan will provide additional confidence that the activation 3.1-12
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 analysis results will provide reasonablejredictions of component activities and resulting dose rates. The efforts are divided into two sections:,
- 1. Verification of the neutron flux predictions by comparison of predicted activities with measurements taken from activated samples of known composition.
- 2. Sensitivity analysis performed on the PCRV concrete trace element abundances to evaluate potential sensitivity of the amount of PCRV concrete to be removed to satisfy final i ' release limits, i
- \
Neutron-Flux Verification: l The validation of the neutron flux will be assessed by comparison of predicted specific activities to measured activities for specimens of known composition. Wireycimens have been removed - from the i PCRV top head and PCRV vertical tendon wires nave been removed from the PCRV side walls to perform this comparison. The comparison of the measured specific activity of the wire removed from the PCRV top head (adjacent to the PCRV liner) and the analytical predictions from the activation analysis indicate that the thermal neutron flux predictions at that location are quite good, within about 4%. The preliminary results of the quantitative analysis indicate that the thermal flux at the tendon location (32 inches -beyond the sidewall PCRV liner) was under-predicted-by the activation analysis by a factor of 2.5 to 3. This increased dose rate would require approximately 4 inches of additional concrete to be removed to obtain a-dose rate in the-5 microR/hr range. This - agreement between predicted and measured thermal flux is
- considered to be good in light of the distance from the core, and 1
the number- of mean- free paths (about 12) traveled by the neutrons through the concrete before reaching the tendon wire. The results of the radi ' emical analysis are being independently verified. Material Composition Verification Analytical sensitivity studies have been performed for the homogeneous rebar/ concrete mixture to assess the effect of trace element abundances on the amount of concrete required for removal in the radial direction. The studies show that if the maximum l 3.1-13
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3-abundances of trace elements identified in Reference 5 are used (rather than the average abundances), an additional 2 inches of concrete would require removal. _In order to estimate the trace element abundances in the PCRV concrete, surface samples of concrete have been taken and_ are being analyzed for their constituents.
Conclusions:
Current data from the top head wire, preliminary data from the vertical tendon wire, and the sensitivity analysis on trace element abundances provide strong evidence that the predictions from the activation analysis are reasonable. In a worst case scenario, where the flux has been underpredicted by a factor of 2.5 to 3 and maximum trace elements are assumed, the depth of concrete required to be removed to meet the free release criteria would increase from 21 inches to approximately 27 inches. Section 2.3.3.12 describes the dismantlement _ plan for removal of the PCRV sidewall concrete. This plan is based on use of the diamond wire saw to cut between every third tendon tube located in the inner row of tendon tubes, located 32 inches from the PCRV sidewall. This cutting technique will remove, as a minimum, a depth of approximately 27 inches pf concrete. This minimum removal depth is adequate to meet the act.ivation analysis requirements _(21 inches) and accommodate maximum uncertainties due to neutron flux (4 inches) and material composition (2 inches). The verification of the vertical tendon wire activity, constituent data from the PCRV concrete samples, and the existing data will allow better estimates of the required depth of concrete to be removed. This will provide additional evidence that the results of the activation analysis are reasonable. Additional details of the verification effort perfermed to date can be foend in " Summary of Existing and Planned Work: for Activation Verification", (Attachment to Reference 7). 3.1.4.2 Plateout Analysis for PCRV Internal Components 3.1.4.2.1 Plateout Analysis Bases and Computer Codes A plateout distribution analysis of radioactive nuclides produced in the reactor core was performed for the PCRV and internal components (Ref. 8). The purpose of this analysis was 'to estimate the plateout concentrations and distributions in the primary coolant circuit. Analyses were conservatively performed from the beginning of cycle (80C) I to the end of cycle (EOC) 5. The axial and radial core 3.1-14
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 l l power distributions through fuel cycle 5 were calculated and used l with flux distribution data as input to fission product release codes. Full-core fuel and graphite temperature distributions, fuel failure and release of key fission gases and metals were then
-calculated. Based on the full-core analysis for key fission gases -and metals, the total plateout and helium purification system inventories of radioactive nuclides were estimated.
Plateout distributions were calculated using the PADLOC (Ref. 9) computer code. The PADLOC code performs a mass transfer calculation l using mass transfer correlations and sorption isotherms to determine ; the partitioning of condensable radionuclides between the flowing l l coolant and the fixed surface in a recirculation loop. The plateout i L model in PADLOC is limited to one-dimensional cylindrical geometry, such that all components of the primary circuit must be modeled as an equivalent series of coupled sections of parallel banks of cylindrical tubes. Reference sorption isotherms were used to ! describe the sorptive capacity of the primary circuit materials for 1 the radionuclides of concern. 1 3.1.4.2.2 Pjateout Analysis Methodoloav
- Typically, the two dominant sources of fission products released from the core are heavy metal contamination (heavy metal outside the coated fuel particles) and fuel particles whose coatings fail in service. In addition, the volatile metals (Cs and Sr) can, at sufficiently high temperatures and over -long periods of time, diffuse through the si! icon carbide (Sic) coatings and be released from the intact fuel particles.
, Calculations were performed for-the following key nuclides: Sr-90, 1-129, 1-131, Cs-137, Cs-134 and Te-127m.- The source terms for fission product plateout analysis include both a direct release contribution and, where applicable, a precursor contribution. In L the case of the cesium isotopes, there is a direct release of both Cs-137 and Cs-134 metal from the core. Cs-137 plateout also results
'from-the release and subsequent decay of its precursor contributor, Xe-137.. Cs-134 has no gaseous precursor. Similarly for Sr-90,
- -there is a direct Sr-90 metal release as well as the contribution l 'from it Kr-90 precursor. Only direct release contributions are p considered fro I-129, 1-131 and Te-127m.
3.1.4.2.3 PlateouL Analysis Results It is anticipated that any internal PCRV component that has.come in contact with primary coolant will require decontamination or will be 3.1-15 l
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 iemoved for disposal as radioactive waste. This includes not only the core graphite and structural components (which are al so activated), but also the steam generator modules, helium circulators and Kaowool insulation. The preliminary results of the plateout analysis are shown in Tables 3.1-4 and 3.1-5. Table 3.1-4 lists the plateout concentration (Ci/cm2) on primary circuit components for the key nuclides, Cs-137 and Sr-90. Table 3.1-5 identifies the integrated plateout (Ci) of primary circuit components for the following nuclides: Cs-134, Cs-137, 1-131, 1-129, Sr-90 and Te-127m. Additional information on the analysis results, analytical models and comparisons with measured data are located in Reference 8. Additional plateout analyses will be performed to predict primary system inventories based on the actual end of life burnup. The accuracy of the predicted fuel performance and gas release will be assessed by comparison to measured R/B (Release to Birth Rate) data. The accuracy of the predicted fission metal release data will be assessed by comparison to measured plateout probe data. Predicted aforementioned assessments to provide best estimates of total plateout inventories. Plateout distributions and concentrations will then be calculated for the primary circuit. This study is scheduled for completion in January 1991. 3.1.4.3 Contaminated Systems. Structures and Components In August 1990, comprehensive radiation and contamination surveys were performed in the Reactor and Turbine Buildings to identify the major contributors to radiation levels above background. Due to on-going maintenance, defueling, and component removal activities in Fogress at the time of the survey, it should be noted that radiation and contamination levels may vary due to the movement of various radioactive components. Additionally, certain PCRV internal components may be removed from the PCRV prior to commencement of decommissioning activities which may change radiation and contamination levels elsewhere in the plant. When compiling the radiation survey results, the contributions from various pieces of portable equipment such as ventilation units, vacuum cleaners, decontamination equipment, etc., were neglected due to the transient nature of their use. An engineering analysis of the total curie inventory at Fort St. Vrain was completed in June 1989 and the results of this analysis have been summarized in Table 3.1-6. This analysis is based upon past survey resuits, activation analysis, plateout analysis and 3.1-16
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 general estimation of contamination levels-occurring in the various systems. The survey results and estimation of contamination levels were then applied over the estimated surface area of the associated system. This analysis accounts for all expected radioactivity at Fort St. Vrain with the exception of fuel. Section 3.1.2 contains a detailed summary of the radiation survey results. These surveys were performed to identify general radiation and contamination levels in frequently accessed areas of the facility. More detailed surveys of individual areas will be required when determining specific work plans during actual decommissioning. 3.1.5 Initial Site Characterization Plans The initial site radiological characterization will be performed to determine the radiological status of Fort St. Vrain balance of plant systems, auxiliary systems, buildings and site within approximately 1000 feet of the Fort St. Vrain facility. Radiological measurements for direct radiation, residual contamination (fixed and removable) will be conducted and recorded. Samples will be taken from strategically selected locations, analyzed and recorded. Results from current Fort St. Vrain radiological data will be used as part of the initial characterization, where appropriate. l ! The results of the initial site characterization will assist in the determination of final survey plans, frequency of surveys and instrumentation to be used. It will also be used as a . general performance indicator to assess the effectiveness of the overall site decontamination. The data will be utilized for radioactive
-waste management, assessing potential hazards during the decontamination and decommissionir.g work, for determining safety controls, and accurately scheduling the decommissioning activities.
I' l^ i l L 3.1-17 1
,ua >,u-~ s.a- unn , - ----a.A--- m .~A-- ---- e &.--- n.. -M --a - --L. _a... s x--a._., - - - -- es- - + ~ . .
PROPOSED DECOMMISSIONING PLAN 6/28/91 l SECTION 3 i INTENTIONALLY LEFT BLANK l l 3.1-10
l PROPOSED DECOMMISSIONING PLAN 6/28/91 TABLE 3.1-2 SECTION 3 ACTIVATION Al.ALYSIS RESULTS l ! (Total Curies Three Years After Shutdown) Total Mn-54 Fe-55 fe-59 Wi-59 Ni-63 tb-w. Ag-110 Eu-152 Eu-154 others curies Fixed Corponents: H-3 C-14 Ca-41 Co-45
<0.01 - - -
0.01 8.43 0.11 7.28 <0.01 -
- 1. Core Barrel -
<0.01 - <0.01 142.29
<0.01 139.96 <0.0% <0.01 0.04 - -
- 2. CSF Liner -
0.C3 <0.- - - <0.01 114.79 0.08 110.34 <0.' <0.01 -
- 3. PCRV Liner -
- 4. CSF/Kaowool Insuta- - <0.01 89.52
<0.01 <0.01 87.20 <0. 4 <0.01 0.08 - - -
tion and Cover Plates - <0.01 -
<0.01 0.22 - - - <0.01 253.09
<0.01 <0.01 - <0.01 246.54 <0.0. ..
- 5. CSF silica Blocks -
- 6. PCRV Kaowool Insuta- - - - 0.01 6.03
<0.01 <0.31 -
0.01 5.57 <0.01 0.44 - - - tion ard Cover Plates -
- 7. Metal Shett-large - - - -
<0.01 0.01
- - <0.01 0.01 <0.01 <0.01 - -
Side Reflector - -
- 8. Large Side Reflector eruf Permanent - - - .21 527216.00 0.30 441114.00 <0.01 3446.24 - -
Hexagonal Blocks 82557.70 - 20.11 77.44
<0.01 116.89 47.19 <0.01 0.01 <0.01 69.15 - 0.54 - - -
- 9. Core Stoport Blocks -
- - - - <0.01 0.01
<0.01 0.01 <0.01 <0.01 - -
- 10. Reflector Keys
- 11. Boronated Spacer - 0.26 66237.70 0.81 3.12 <0.01 47208.60 :0.01 7097.13 2.81 392.45 - -
Blocks 11531.50 1.02 594.L2 TOTAL Demovable Corponents: 290.39 0.13 - - 0.01 230C7.20
- - - 0.06 15451.70 <0.01 4342.78 2.08 -
- 1. Metal Clad Block-CR -
<0.01 - - -
<0.01 E 55.01 0.52 169668.20 <0.01 2786.29
- 2. Metal Olad Block-NCat 0.01 - <0.01 122.14
<0.01 0.07 71.85 <0.01 48.93 0.01 1.27 - -
- 3. Region constraint Device - - -
- - - - <0.01 414.98
- - 0.17 299.25 <0.01 115.56 - -
- 4. Orifice Valve - -
0.67 88.35 <0.01 - - - <0.01 3799.40 1.38 -
<0.01 0.07 301.45 <0.01 3407.48
- 5. Reflecter Block e 8th 3
Hastelloy Cans 199.878.73 TOTAL
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6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 TABLE 3.1-5 INTEGRATED PLATE 0VT IN EACH PRIMARY CIRCUIT COMPONENT AT E0C5 BRANCH Cs-134** Cs-137** I-131* I-129* Sr-90* Te-127m* , NAME (Curies) (Curies) (Curies) (Curies) (Curies) (Curies) Active Core 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Lower Reflector 1.510E+01 1.326E+01 6.077E-03 1.037E-08 5.820E-01 3.196E+01 Core Support Blocks 3.668E-01 3.224E-01 2.923E-04 4.987E-10 1.02CE-02 5.664E-01 Core Exit Plenum 8.166E-02 7.195E-02 2.028E-04 3.461E-10' l.937E-03 1.000E-01 Steam Generator Inlet 1.617E-03 1.443E-03 9.125E-03 1.557E-08 1.7703-02 8.049E-01 Steam Gentrator Reheater 1.029E+01 8.771E-01 6.272E-01 1.066E-06 4.527E-01 1.923E+01 Superheater 2.699E+00 2.305E+00 1.478E+00 2.516E-06 1.544E-01 6.909E+00 Economizer 7.513E+00 6.339E400 2.788E+01 4.341E-05 2.244E-02 9.067E-01 Evaporator 7.398E+00 6.879E+00 1.120E+03 4.868E-04 8.131E-03 1.696E-01 Steam Gen. Outlet Plenum 2.277E-02 2.272E-02 1.164E401 2.531E-05 6.615E-04 4.924E-04 ' Circulators 1.458E-01 1.388E-01 4.432E+00 7.501E-06 8.754E-03 3.305E-03 Circulator Outlet Plenum 5.385E-03 6.882E-03 6.929E+00 2.5073-05 1.274E-03 1.161E-04 Core Barrel / Liner Annulus 1.196E-01 1.886E-01 4.172E+01 8.176E-05 5.435E-02 2.579E-03 Core Inlet Plenum 1.410E-02 3.302E-02 1.943E+01 7.128E-05 1.315E-02 3.041E-04 Upper Reflectors 8.109E-01 2.448E+00 2.702E-01 6.322E-07 1.069E+00 1.752E-02 Side Reflectors 1.053E-03 5.767E-03 1.761E-01 5.441E-07 1.838E-03 '2.210E-05 Purification System 1.106E-03 2.902E-03 2.853E+00 3.542E-04 1.293E-03 2.404E-05 TOTAL (E0CS) 3.530E+01 3.290E+01 1.237E+03 1.100E-03 2.400E+00 6.067E+01 TOTAL (3 YEARS DECAY) 1.290E+01 3.071E+01 - 1.1003-03 2.234E+00 5.739E-02
- Based upon the source rate calculated from the xenon data using the square root of half-life dependence.
**Plateout distribution based upon sorption isotherms for unoxidized alloy steel surfaces.
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3: TABLE 3.1-6 ESTIMATED CURIE TOTAL AT FSV (Three Years After Shutdown) NOTE: The systems listed below are those systems which are knownj to be contaminated, on-going maintenance, defueling and component removal may transfer contamination to other systems and/or locations. Total Curies System From From Loose No . - System Activation Contuaination
- 11' ' PCRV and Internal 7.94 E+05 2.54 E+02 Components 12 Controls Rods and Drives 1.84 E+04 N/A 13 Fuel Handling Equipment N/A 8.95 E-03 14 Fuel. Storage Facility- N/A 2.08 E-02 16- Auxiliary Equipment N/A 9.05 E-03 17 Reactor Removable Reflector 4.82 E+05 N/A
.21 Primary Coolant- N/A 6.01 E+01 22 Secnndary Coolant N/A 5.68 E403 23 Helium Purification N/A 9.33 E-01 61 Decontamination Systems' N/A 1.06 E 05-62- Radioactive Liquid Waste- N/A 4.06 E-05 63 Radioactive Gas Waste N/A 8.15 E-05 Includes an estimate of loose surface contamination due to activated corrosion products.
G
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i l L 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 L 342 RADIATION PROTECTION PROGRAM i
- 3.2.1 Introduction i This section sets fortie the policy and the requirements of the j Radiation Protection Program to be implemented during the l
- decommissioning of the Fort St. Vrain facility. This section is the R highest tier document of the Radiation Protection Program and provides definitions of the Radiation Protection. organization, responsibilities, authorities, administrative policies,' program objectives and standards to implement the Radiation Protection
. - Program. This section will also be used as the basis document for [ all Radiation Protection Program administrative and implementing procedures. Title 10 Code of Federal Regulations, applicable regulatory guidance
' documents and industry standards are the basis of the Radiation Protection - Program. This section was specifically formatted using Draft NUREG-0761, " Radiation Protection Plans for Nuclear Power Reactor Licensees"-(Ref. 10), which provides guidance for the l
content of- a " Radiation Protection Plan". It.also incorporates the guidance contained in - NRC Regulatory Guide 8.8, "Information
- Relevant to- Ensuring that Occupational Radiation Exposures at Nuclear Power- Stations _will 'be as Low as is Reasonably Achievable"
_(Ref,11) and' NRC Regulatory Guide- 8.10, " Operating Philosophy for Maintaining Occupational-Radiation Exposures as Low as is Reasonably
- Achievable" -(Ref. 12). The Radiation Protection Program will-incorporate the requirements of the 1991 revision to 10 CFR 20 prior to thenstart of any physical decommissioning work..
3.2.2c ~Manaaement Polict l~ 3.2.2.1 JManaaement Policy Statement LPSC and its. management are committed to the safe _ decommissioning,0f the Fort St. Vrain facility. The primary objective of'the Radiation Protection Program is to protect the work.ers, -visitors and- the - general _ public from radiological hazards that have the potential of developing during' the decommissioning project. . PSC _and its contractors will provide sufficient qualified staff, facilities and' equipment to perform _ the Fort St. Vrain facility decommissioning in a radiologically safe manner. PSC~is comcitted to strict compliance
-.with regulatory requirements, radiation exposure limits, and limits regarding rele'ase of radioactive materials. In addition, PSC will make every _ reasonable effort to maintain radiation exposures and-releases of radioactive-materials in effluents to unrestricted areas 3.2-1 i
~ g ,m- e w w y
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 "As low As Reasonably is Achievable" (ALARA). The ALARA philosophy will be incorporated into all decommissioning activities and have full management support. 3.2.2.2 Administration Policy Activities conducted during the Fort St. Vrain decommissioning project that have the potential for exposure to radiation or radioactive materials will be managed by qualified individuals who will perform program operations according to procedural guidelines. Radiological hazards will be monitored and evaluated on a routine basis to maintain radiation exposures and the release of radioactive materials to unrestricted areas as far below specified limits as is reasonably achievable. All decommissioning project work activities, and each element of the Radiation Protection Program will be specifically defined and implemented using written manuals, procedures and instructions. Radiation protection training will be provided to all occupationally exposed individuals to ensure they understand and accept the responsibility to follow all procedures and to maintain their radiation dose ALARA. Project management will ensure that work specifications, designs, and work packages involving potential radiation exposure or handling of radioactive materials incorporate effective radiological controls. Project supervisors will include radiation protection considerations in the work activities under their control. Radiation protection records will be prepared and maintained using high standards of accuracy, traceability and legibility to meet the requirements of regulatory agencies and company procedures. 3.2.2.3 ALARA Policy l- All activities at Fort St. Vrain involving radiation and radioactive materials shall be conducted such that radiation exposures to employees, contractors, and the general public are maintained ALARA, taking into account current technology and the economics of radiation exposure reduction in relationship to the benefits to l health and safety. Project management will establish specific goals and objectives for the Fort St. Vrain decommissioning project ALARA program. The ALARA l program will be based on the guidance provided in Regulatory Guides 8.8 and 8.10 (References 11 and 12). The ALARA program will incorporate current technology and sound radiation protection practices to maintain exposure to ionizing radiation ALARA. 3.2-2
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 3.2.2.4 Reaulatory Comoliance Policy Project management will maintain the Radiation Protection Program in compliance with the requirements of 10 CFR (as amended), 49 CFR and to the extent practical, the information contained in industry standards, Regulatory Guides and other guidance documents referenced in Section 3.2. The decommissioning of Fort St. Vrain is scheduled to occur during the transition to the revised 10 CFR 20 regulations. For this reason, PSC is committed to implement all of the provisions of the revised 10 CFR 20 prior to the start of any physical decommissioning work. Radiation protection procedures will be prepared using the most current regulatory guidance available. Following implementation, the Radiation Protection Program will be periodically assessed against any newly issued regulatory guidance and modified, if necessary. 3.2.2.5 Waste Minimization and Disposal Policy Project management will implement and enforce a program for minimizing the generation of radioactive wastes. Implementing procedures will be developed for the use, classification, treatment, packaging and shipment of radioactive material. These procedures will ensure strict compliance with applicable Federal, State and local regulation and burial site criteria. Project management will establish waste minimization goals. To ensure these goals are achieved, all decommissioning personnel will receive training in the applicab? e procedures and practices to minimize the generation of radioactive waste. 3.2.2.6 Respiratory Protection Policy Project management is committed to minimizing the inhalation of air contaminated with dusts, mists, fumes, gases, vapors and radionuclides. The primary means of achieving this goal will be to prevent or mitigate the hazardous condition at the source. Every reasonable effort will be made to achieve this objective by ~using engineering controls, including process modification, containment and ventilation techniques. The use of respiratory protection equipment will be consistent with the goal of maintaining the total effective dose to personnel ALARA. A respiratory protection program will be developed, implemented and maintained in accordance with 10 CFR 20 and using the regulatory guidance found in NRC Regulatory Guide 8.15, " Acceptable Programs for Respiratory Protection" (Ref.13), and the NUREG-0041 " Manual of 3.2-3
PROPOSED DECOMMISSIONING PLAN 6/2fj/91 SECTION 3 Respiratory Protection Against Airborne Radioactivity Materials" (Re f. 14 ) . 3.2.3 Ibdiation Protection 0.tganization and functions 3.2.3.1 Radiation Protection Oroanization The Radiation Pr^tection Organization will ensure that a high level of performance in radiation protection is achieved through effective implementation and control of radiation protection activities. This high level of radiation protection performance will be achieved ihrough the combined efforts of PSC management and the Westinghouse Team. The PSC management structure that will oversee and control the Radiation Protection Program during the decommissioning project is shown on figure 2.4-1, "PSC Decommissioning Organization Chart". The PSC organization will provide control, direction and oversight, and en ure the implementation of the Radiation Protection Program. The PSC Program Manager for Decommissioning and the Westinghouse > Project Director have ultimate responsibility for assuring that an effective Radiation Protection Program is implemented during the fort St. Vrain decommissioning. This corporate and project interface will ensure a coordinated and effective approach to the minimization of individual and collective dose and the control of radioactive materials during decommissioning. Reporting directly to the PSC Program Manager for Decommissioning will be the PSC Facility Support Manager. The PSC Facility Support Manager- has . oversight responsibility for the development and implementatica of the Radiation Protection Program policies and standards and is the PSC Radiation Protection Manager (RPM) for the decommis?ioning project. The PSC Facility support Manager will serve as a member of the Decommissioning Safety Review Committee and will also serve as the Chairman of the ALARA Committee. Tha PSC Facility Support Manager will ensure that PSC has the proper control and authority over decommissioning activities as they relate to radiation protection. The PSC Facility Support Manager represents the formal line of communication and authority between Fort St. Vrain management and the Westinghouse organization for radiation protection matters. The PSC Facility Support Manager will be responsible for approval of the Radiation Protection Program manuals and the content of. radiation protection training programs. The PSC facility Support Manager _will also be directly responsible for the Radiological Environmental Monitoring Program and the Emergency 3.2-4
6/28/91 PROPOSED DECOMMISSIONING PLAN SECil0N 3 Response Plan. The PSC Ficility Support Manager will be qualified in accordance with NRC Regulatory Guide 1.8 " Personnel Selection and Training" (Ref. 15), and ANS/nNSI 3.1 " Selection, Training and Qualification of Personnel for Nuclear Power Plants" (Ref.16). The staff positions (llealth Physicists) reporting to the PSC Facility Support Manager will provide review and evaluation functions to ensure that the Rr.diation Protection Program policies and standards are implemented. The Westinghouse Project Radiation Protection Organization will be administered by the Project Radiation Protection Manager (PRPM), under the authority of the PSC Radiation Protection Manager. Figure 3.2-1, " Westinghouse Team Radiation Protection Organization Chart" shows the key membeis of the Radiation Protection Organization. The Project Radiation Protection Manager, under the direction of the PSC Facility Support Manager, has the responsibility for the Rhdiation Protection Program development, implementation and compliance with the applicable regulations. The Project Radiation Protection Manager will be qualified in accordance with NRC Regulatory Guide 1.8 (Ref. 15) and ANS/ ANSI 3.1 (Ref. 16). The Project Radiation Protection Manager will report directly to the Westinghouse Project Director. This reporting chain will ensure sufficient authority and independence to implement an effective Radiation Protection Program. It will also provide a direct line of communication to senior project r:nagement. The Project Radiation Protection Manager will have the authority to stop work whenever activities have the potei.tial to jeopardize the health and safety of workers, visitors or tLe general public. This authority will not be limited to radiological safety issues. If the activities violate operational parameters, administrative guidelines, safety requirements or Radiation Protection procedures, the Project Radiation Protection N, nager will have the authority to stop work. The authority to overrule the Project Radiation Protection Manager's stop work order may only come from the PSC Program Manager for Decommissioning, PSC Facility Support Manager or the Westinghouse Project Director. The staff positions shown on Figure 3.2-1 will have the primary responsibility for providing technical direction, implementation of the Radiation Protection Program, and supervision of the activities of the Radiation Protection Technicians and support personnel. Designated radiation protection staff members will be qualified in accordance with NRC Regulatory Guide 1.8 (Ref. 15) and ANS/ ANSI 3.1 (Ref. 16), and will serve as the qualified substitute for the 3.2-5
PROPOSED DECOMMISSIONING PtAN 6/28/91 SECTION 3 Project Radiation Protection Manage- The staff positions will have the authority to stop work whenever activities jeopardize the radiological health and safety of workers, visitors or the general public, or, if the activities violate Radiation Protection procedures. The number and titles of positions shown on Figure 3.2-1 may be modified during the course of decommissioning. This may be necessary during initial project start-up and demobilization. Changes to the Radiation Protection organization will require approval from the PSC Facility Support Manager (PSC Radiation Protection Manager). 3.2.3.2 Bintional Descripligm The effective implementation of the Radiation Protection Program is the responsibility of all project personnel. Specific responsibilities for the implementation of the Radiation Protection Prooram are listed below. The PSC Vice President, Nuclear Operations, is responsible for the safe decommissioning of Fort St. Vrain and is the designated Corporate Officer for PSC. The PSC Program Manager for Decommissioning is responsible for conducting facility decommissioning in accordance with regulatory requirements including those activities related to radioactive materials and radiation exposure. Major responsibilities related to the Radiation Protection Program include the following:
- 1. Ensure support for the ALARA program from project personnel.
- 2. Participate in the selection of specific radiation protection goals and objectives for the decommissioning.
- 3. Support the Radiation Protection Manager in implementing the Radiation Protection Program.
- 4. Ensure periodic status reports on the Radiation Protection Program are distributed to management.
- 5. Issue or rescind "stop work" orders, as required.
- 6. Oversee the Decommissioning Emargency Response Plan.
The PSC Facility Support Manager reports to the PSC Program Manager for Decommissioning and is responsible for implementatior, of labnratory, environmental monitoring, training, access control, emergency response and Radiation Protection Program activities. l 3.2-6 '
-. =-
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 Major responsibilities related to the Radiation Protection Program include the following:
- 1. Ensure proper impicmentation of Fort St. Vrain Radiation Protection policy. ,
- 2. Interface with the Project Radiation Protection Manager. l
- 3. Ensure adequate staffing, facilities and equipment are i available to perform the functions assigned to Radiation Protection personnel.
- 4. Serve as the PSC Radiation Protection Manager responsible for oversight and direction of the Radiation Protection Program.
- 5. Issue or rescind "stop work" orders, as appropriate.
- 6. Ensure personnel at Fort St. Vrain have received job specific and general employee training.
- 7. Serve on the Decommissioning Safety Review Committee.
- 8. Serve as chairman of the ALARA Committee.
- 9. Ensure proper disposal of radioactive solid, liquid and gaseous wastes.
- 10. Maintain and implement the Decommissioning Emergency Response Plan,
- 11. Coordinate revisions to the Radiation Protection Program.
- 12. Approve Radiation Protection training programs.
- 13. Ensure implementation of the Radiological Environmental Monitoring Program.
- 14. Review and approve training programs related to work in radiological areas or involving radioactive material.
The PSC Health Physicists report to the PSC Facility Support Manager. Major responsibilities as related to the Radiation Protection Program include the following:
- 1. Coordinate the annual review of the Radiation Protection Program.
- 2. Serve as designated alternates to the PSC Radiation Protection llanager.
- 3. Evaluate Radiation Protection training programs.
- 4. Monitor collective exposure of various decommissioning activities.
- 5. Conduct inspections of work in progress to evaluate the adequacy of the implementation of the Radiation Protection Program.
- 6. Evaluate plant contamination control activities.
- 7. Support the decommissioning ALARA committee.
- 8. Review radiological occurrences to identify root cause and corrective actions for radiation protection incidents.
3.2-7 L l
~-c- m - - - , , ---r,, , ,-r ,, -n n ,w- -,y -r
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3
- 9. Issue and rescind "stop work" orders, as appropriate.
- 10. Evaluate the adequacy of the radioactive material disposal and shipping activities.
- 11. Evaluate the dosimetry and personnel exposure tracking system.
- 12. Evaluate bioassay and radiochemistry activities.
- 13. Evaluate the calibration of portable, stationary and laboratory radiation monitoring equipment.
- 14. Evaluate the adequacy of the Respiratory Protection Program.
- 15. Evaluate the Radiological Environmental Monitoring Program.
- 16. Participate in the Decommissioning Emergency Response Plan.
The PSC Project Assurance and Operations Managers' responsibilities as related to the Radiation Protection Program include the following:
- 1. Ensure personnel under their direction are properly trained in and comply with radiation protection requirements.
- 2. Support the PSC Radiation Protection Manager in overseeing the implementation of the Radiation Protection Program, including (but not limited to) performance of audits and surveillances of their areas of responsibility, and routine inspections of work areas where their personnel are involved.
- 3. Participate in the Decommissioning Emergency Response Plan.
The FSC Shift Supervisor reports to the PSC Operations Manager. H&jor respcnsibilities as related to the Radiation Protection Program include the following:
- 1. Ensure the safe operation of plant systems.
- 2. Ensure planned radiological effluent releases are properly performed.
- 3. Notify the Nuclear Regulatory Commission, as required.
- 4. Notify Radiation Protection personnel when changes in plant conditions could affect radiological conditions.
- 5. Ensure personnel under their direction comply with the requirements of the Radiation Protection Program.
- 6. Participate in the Decommissioning Emergency Response Plan.
3.2-8
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 1he Westinghouse Project Director is responsible for conducting decommissioning operations in accordance with regulatory requirements related to radioactive materials and radiation exposure. Major responsibilities as related to the Radiation Protection Program include the following:
- 1. Ensure that all project personnel are properly trained in and comply with radiation protection requirements.
- 2. Ensure support for the ALARA program from all project personnel.
- 3. Participate in the selection of specific Radiation Protection goals and objectives for the decommissioning.
- 4. Support the Project Radiation Protection Manager in implementing the Radiation Protection Program.
- 5. Ensure periodic status reports on the Radiation Protection Program are distributed to project management.
- 6. Issue or rescind "stop work" orders, as appropriate.
- 7. Participste in the Decommissioning Emergency Response Plan.
The Westinghouse Site Operations Manager, Technical Services Manager and Project Control Manager report directly to the Westinghouse Project Director. Major responsibilities as related to the Radiation Protection Program include the following:
- 1. Ensure personnel under their direction are properly trained and comply with radiation protection requirements.
- 2. Support the Radiation Protection Manager in the implementation of the Radiation Protection Program.
- 3. Ensure that exposure and waste reduction techniques are incorporated into work plans and procedures.
- 4. Ensure radiation protection and ALARA principles are incorporated into project activities.
- 5. Ensure that requirements, methods, regulations and procedures for waste processing are specified.
- 6. Ensure that craft labor is provided for operation of the waste processing system, waste segregation and packaging.
- 7. Participate in the Decommissioning Plan.
The Westinghouse Project Radiation Protection Manager reports directly to the Westinghouse Project Director and is responsible for laboratory analysis, radiation protection training, radioactive waste management, dosimetry, respiratory protection, ALARA and radiation protection job coverage. Major responsibilities related to the Radiation Protection Program include the following: 3.2-9
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3
- 1. Ensure the implementation of the Radiation Protection Progr?m policies and procedures.
- 2. Review and approve Radiation Protection procedures.
- 3. Provide an interface for the PSC Radiation Protection Manager and the Westinghouse Project Director.
- 4. Ensure adequate staffing, facilities and equipment are available to perform the functions assigned to Radiation Protection personnel.
- 5. Select and approve Radiation Protection staff members.
- 6. Ensure that adequate radiation protection coverage (minimum established ratio of workers to Radiation Protection Technicians) is provided for project personnel during all working hours.
- 7. Establish goals and objectives for the Radiation Protection Program.
- 8. Issue or rescind "stop work" orders, as appropriate.
- 9. Ensure that locations, operations, and conditions that have the potential for causing significant exposures to radiation are identified.
- 10. Review and approve training programs related to work in radiological areas or involving radioactive material,
- 11. Provide training for personnel to work at fort St. Vrain including job specific General Employee Training.
- 12. Ensure proper disposal of radioactive solid, liquid, and gaseous wastes.
- 13. Provide Radiation Protection input to decommissioning planning:
- 14. Trend radiation work performance of project personnel including contamination and radiation exposure control.
15.. Identify and review causes and corrective actions for incidents associated with radiation protection.
- 16. Ensure an effective ALARA Program.
- 17. Serve as member of the ALARA Committee.
- 18. Participate -in the Decommissioning Emergency Response Plan.
The Westinghouse Project Radwaste Supervisor reports to the Project Radiation Protection Manager. Major responsibilities as related to the Radiation Protection Program include the following:
- 1. Coordinate radioactive waste minimization and. disposal activities.
- 2. Classify radioactive waste material in accordance with 10 CFR 61 in preparation for disposal.
- 3. Monitor the packaging and preparation of radioactive material for shipment.
3.2-10 L
6/28/91 PROPOSED DECOMMISSIONING PLAli SECTION 3
- 4. Schedule and complete shipments of radioactive material in accordance with Department of Transportation (DOT) and Ni.C rei, lations.
- 5. Make recommendations to the Project Radiation Protection Manager concerning site management issues that affect radwaste operations and shipping.
- 6. Provide trairing for radioactive waste management.
- 7. Participate in the Decommissioning Emergency Responso Plan.
The Westinghouse ALARA Supervisor reports to the Project Radiation Protection Manager. Major responsibilities as related to the Radiation Protection Program include the following:
- 1. Review of work packages for ALARA consideration.
- 2. Establish ALARA person Rem budgets for project tasks.
- 3. Coordinate pre-job briefings and mock-up training.
- 4. Track project exposure and prepare project person Rem reports.
- 5. Serve as ALARA committee secretary.
- 6. Implement the ALARA suggestion program.
- 7. Coordinate the ALARA exposure history records management system.
- 8. Perform cost benefit analyses, as required.
- 9. Provide radiation protection training, as required.
- 10. Participate in the Decommissioning Emergency Response Plan.
The Westinghouse Radiation Protection Operations Supervisor reports to the Project Radiation Protection Manager. Major responsibilities as related to the Radiation Protection Program include the following:
- 1. Ensure that policies relating to personnel radiation exposures are established and enforced.
- 2. Ensure appropriate radiation and contamination surveys are performed to verify radiation levels and working conditions of the facility.
- 3. Ensure accountability of all byproduct material.
- 4. Train technicians to handle all phases of radiation protection work.
- 5. Review program effectiveness.
- 6. Issue and rescind "stop work" orders, as appropriate.
- 7. Coordinate the Respiratory Protection Program.
- 8. Control portable, stationary and laboratory radiation monitoring equipment.
1 3.2-11
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3
- 9. Interface with the Westinghouse training group.
- 10. Prepare procedures for Radiation Protection operations. ,
- 11. Review dosimetry and bioassay results.
- 12. Act as ALARA Supervisor, as requested.
- 13. Ensure implementation of the Radiation Work Permit Program.
- 14. Participate in the Decommissioning Emergency Response Plan.
The Westinghouse Technical Support Supervisor reports to tte Project Radiatinn Protection Manager. Major responsibilities as related to the Radiation Protection Program include the following:
- 1. Ensure araropriate bioassay samples are taken and results reviewed at established intervals.
- 2. Ensure records of personnel exposures are maintained. l
- 3. Ensure radiochemical analyses of solid, liquid, and gaseous samples taken from plant systems and plant environs are performed.
- 4. Maintain proper reagent preparation and control.
- 5. Ensure laboratory quality control.
- 6. Provide radiation protection training support.
- 7. Prepare procedures for radiochemistry operations.
- 8. Ensure proper calibration of stationary, portable and laboratory radiation monitoring equipment.
- 9. Assist Project Radiation Protection Manager in the review of program effectiveness.
- 10. Provide technical direction and oversight in the areas of radiological engineering, respiratory protection, Radiation Protection instrumentation and other areas of the Radiation Protection organization, as necessary,
- 11. Evaluate internal and external dosimetry results.
I '2 . Evaluate programmatic deficiencies and prescribe corrective actions.
- 13. Participate in the Decommissioning Emergency Response Plan.
The Westinghouse Radiation Protection Technicians report to a designated Radiation Protection Supervisor. Major responsibilities related to the Radiation Protection Program include the following:
- 1. Conduct radiation and contamination surveys and keep legible records.
- 2. Identify and post radiation, contamination, hot particle, airborne and radioactive material areas, i 3.2-12
6/28/91 PROPOSED DECOMMISSIONING PLAN SEC110N 3
- 3. Prepare Radiation Work Permits to control access to and activities in radiologically controlled areas.
- 4. Monitor work to assure compliance with good radiological work practices.
- 5. Implement ALARA program requirements.
- 6. Maintain and calibrate portable monitoring instruments.
- 7. Issue "stop work" orders as appropriate.
- 8. Sample various process streams for radiochemical analysis.
- 9. Verify packaging of radioactive material.
- 10. Participate in emergency response activities as delineated in the approved Decommissioning Emergency Response Plan.
All Project Supervisors have major responsibilities related to the Radiation Protection Program including the following:
- 1. Ensure that personnel assigned to work with radioactive material attend required training.
- 2. Ensure personnel under their direction comply with radiation protection requirements.
- 3. Identify radiation work procedures and practices that need upgrading.
- 4. Assign tasks and ensure that workers are prepared for tasks in order to maintain doses ALARA.
- 5. Ensure that employees know the radiological hazards of their duties.
- 6. Ensure that assigned equipment and facilities are designed, installed and operated to minimize the radiological hazards to personnel.
- 7. Know the location and the radiological hazards in the work area.
- 8. Know the exposure status of those for whom they are responsible.
- 9. Provide information on projected work activities to the Radiation Protection organization.
- 10. Notify Radiation Protection personnel of any radiological problems encountered,
- 11. Assign tasks to distribute dose among exposed personnel to minimize the likelihood of overexposures and to maintain individual doses ALARA.
- 12. Ensure that workers are prepared for tasks with tools, equipment and training to minimize time spent in Radiation Areas.
All Project Workers have major responsibilities related to the Radiation Protection Program including the following: 3.2-13
PROPOSED DECOMMISSIONING PL AN 6/28/91 SECTION 3
- 1. Obey promptly "stop work" and " evacuate" orders from Radiation Protection personnel.
- 2. Obey posted, oral and written radiological control instructions and procedures, including instructions on RadiationWorkPermits(RWPs).
- 3. Wear ILDs and self reading dosimeters where required by postings or as directed by Radiation Protection personnel.
- 4. Immediately report unexpected exposure and lost or offscale dosimeter to Radiation Protection personnel. .
- 5. Keep track of personal radiation exposure status to ensure that administrative dose limits are not exceeded.
- 6. Remain in as low a radiation area as practicable to accomplish work.
- 7. Do not loiter in radiation areas.
- 8. Do not smoke, eat, drink or chew in radiologically controlled areas.
- 9. Wear anti-contamination clothing and respiratory protection properly wherever required by postings or by Radiation Protection personnel. '
- 10. Remove anti contamination clothing and respiratory protection properly to minimize spread of contamination,
- 11. Monitor for contamination when leaving a contaminated area or a radiologically controlled area and notify Radiation >
Protection personnel if contamination is found.
- 12. Minimize the spread of contamination and promptly notify ,
Radiation Protection personnel of any known or potential radioactive spills.
- 13. Do not unnecessarily touch a contaminated surface or allow clothing, tools or other equipment to do so.
- 14. Place contaminated tools, equipment and solid waste on disposable surfaces (e.g., sheet plastic) when not in use and inside plastic bags when work is finished.
- 15. Limit the amount of material that has to be decontaminated '
or disposed of as radioactive waste.
- 16. Notify Radiation Protection personnel of faulty or alarming Radiation Protection equipment.
- 17. Report the presence of open wounds to Radiation Protection ,
personnel prior to working in areas where radioactive l contamination exists and exit immediately if a wound occurs while in such an area.
- 18. Notify Radiation Protection personnel upon returning to the site after medical administration of radiopharmaceuticals.
- 19. . Assure a mentally alert and physically sound condition for performing assigned work.
3.2-14
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3
- 20. Ensure that work activities do not create radiologica, problems for others and be alert fer possibilities that activities of others may change the radiological conditions to which the individual is exposed.
- 21. Do not touch or pickup material in RCA's without a prior survey.
- 22. Comply with the requirements of the Decommissioning Emergency Response Plan.
3.2.3.3 Radiation Protection Oraanization Staffing The Radiation Protection organization will provide appropriate personnel and resources to verify a radiologically safe working environment. A sufficient number of Radiation Protection personnel will be present during the various decommissioning activities to ) l ensurn compliance with the Radiation Protection Program and l implementing procedures. Projected Radiation Protection staffing , levels are shown in figure 3.2-1. The Project Radiation Protection ! Hanager will establish guidelines for adequate Radiatien Protection staffing based on radiological parameters and work scope. ' Radiation Protection staffing levels will be periodically reviewed by the PSC and Project Radiation Protection Manager and the Radiation Protection Supervisors, as applicable, to ensure that adequate staffing levels are maintained consistent with current and planned decommissioning activities. Specific work packages will be addressed on a case-by case basis (e.g., by the Radiation Protection Operations Supervisor) to ensure adequate Radiation Protection Technician to worker ratios based on the guidelines provided by the Project Radiation Protection Manager. Contingencies will be in place to adjust staffing levels during the project for routine functions and unanticipated radiological events. These staffing contingencies will ensure that all work is performed in a radiologically safe and timely manner. Staff adjustments will be implemented when needed, but only after review and recommendation of the PSC and Project Radiation Protection Managcrs and approval of
-the Westinghouse Project Director.
Continuous Radiation Protection coverage will be provioed for decommissioning work activities that involve significant radiological huards, such as the removal of unshielded, highly activated / contaminated components from the PCRV. For example, two Radiation Protection technicians would normally be assigned to the PCRV area during component removal and handling. Another example is that additional Radiation Protection technicians may be provided 3.2-15
l l l PROPOSED DECOMMISSIONING PLAN 6/28/91 l SECTION 3 during PCRV concrete cutting operations, which are expected to run 24 hours a day. I intermittent Radiation Protection coverage will be provided to decommissioning work activities that have a minimal potential for significant personnel exposure. The removal of balance of plant systems is an example of a work activity that will be monitored I periodically by Radiation Protection Technicians. 3.2.3.4 fladiation Protection Proaram Manuals The Radiation Protection Program will be integrated into all applicable decommissioning work activities. The Radiation Protection Program will be specifically defined and implemented using a program manual consisting of both administrative procedures and specific implementing procedures. i The Radiatinn Protection Program will incorporate four manuals: ; l
- 1. Radiation Protection Manual
- 2. Radioactive Waste Manual
- 3. Radiation Protection Training Manual
- 4. Off-Site Dose Calculation Manual L
The four (4) manuals will contain the administrative and implementing procedures, which will specify the standards and controls and define corporate and site objectives for the programs described in each manual. The Project Radiation Protection Manager will have the responsibility for the development and implementation of these manuals, following approval by the PSC Project Radiation Protection Manager. The development and control of Radiation Protection procedures will be in accordance with the Quality Assurance Plan (Section 7 of this plan) and will incorporate the following procedural guidelines:
- 1. Clearly defined scope, applicability, limiting conditions and precautions.
- 2. Uniform procedure identification and status (titling- or numbering, location, and- status for page and revision identification).
- 3. Consistent format (for organization, instruction step format, instruction step designation, caution and note format, and page format.
l 3.2-16 l \. - . . _ - , .-. - . , -
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3
- 4. Clearly understood text, using standard grammar, nomenclature and punctuationi concise instruction steps in a logical sequence.
- 5. "lloid points" for procedures with unique and/or high personnel risk.
- 6. Effective grouping of procedures and a clear table of contents for the procedure binder or manual to allow easy location of a particular procedure.
- 7. Review, approval, and issuance of temporary changes and permanent revisions.
- 8. Periodic review of procedures.
- 9. Controls to make procedure use as convenient as possible and to ensure that only approvea copies are available.
figure 3.2 2 , " Radiation Protection Program Manual Structure", shows the hierarchy and organization of the Radiation Protection Program manuals and associated program elements. 3.2.3.5 Centracted Radiation Protection Services Procurement of contracted Radiation Protection services will be provided in accordance with Section 7, the Quality Assurance Plan, the Radiation Protection Program, and bid specifications developed for the decommissioning project. Westinghouse has been contracted to provide the operational radiation protection, radioactive waste management and final site release survey, either directly or. through subcontractors. Examples of subcontracted services include: external dosimetry processing, primary instrument calibration and 10 CFR 61 program sample analyses. 3.2.4 Radiation Protection Trainina and Oualification 3.2.4.1 General Considerations All decommissioning project workers will be provided instruction in radiation protection concepts . commensurate with the radiological hazards they may encounter during the fort St. Vrain decommissioning project. This training is recognized as essential in achieving high standards of performance in radiation protection. The Project Radiation Protection initial training, qualification-and retraining programs will be developed using applicable- guidance contained in NUREG-0761 (Ref. 10), NRC Regulatory Guide- 8.27,
" Radiation Protection Training for Personnel At Light-Water Cooled 3.2-17
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 l Nuclear Power Plants", March 1981 (Ref. 17), HRC Regulatory Guide 8.13, " Instruction Concerning Pre natal Ridiation Exposure", (Ref.
- 18) November 1975 and NRC Regulatory Guide 0.29, ' Inst ruction Concerning Risks from Occupational Radiation Exposure", (Ref. 19)
July 1981. Guidance from these documents will be incorporated into a Radiation Protection Training Manual, which will include both administrative and implementing procedures. Radiation protection training will be provided to three basic work i groups: Non Radiation Workers, Radiation Workers and Radiation ) Protection Personnel. Training for each work group will be i organized as follows: I Non Radiation Workers will receive, as appropriate: l l
- 1. Introduction to Radiation Protection !
- 2. Non radiation w9rker indoctrination Radiation Workers will receive, as appropriate:
- 1. Radiation worker training
- 2. Specialized ALARA training
- 3. Respiratory protection training Radiation Protection Personnel will receive as appropriate:
- 1. Radiation Protection technician training
- 2. Radiation Protection support staff training
- 3. Radiation Protection supervisor training
- 4. Radiation Protection Hanager training
- 5. Radioactive waste management training All classroom training will be conducted using lesson plans approved by -the Project Radiation Protection Manager and the PSC facility Support Manager. On-the-job training (0JT) will be administered by a qualification card or equivalent which is approved by the Project Radiation Protection Hanager. Personnel assigned to perform Radiation Protection training will be qualified as instructors and/or evaluators in accordance with the Radiation Protection Training Manual.
3.2.4.2 httroduction to Radiation Protection All project personnel, visitors and transients granted unescorted access to the Restricted Area will receive, as a minimum, annual instruction in elementary radiation ef fects and basic aspects of 3.2-18 i L _ _ . _ _ _
6/20/91 PROPOSLO DECOMMISSIONING PLAN SEC110N 3 l radiation protection during general employee orientation. This training will meet the requirements of 10 CFR 19. . 1 The Radiation Protection orientation training, which will normally be given during General Employee Training, will include:
- 1. Instructions to not enter radiologically controlled areas, violate radiological postings or cross radiological boundaries.
- 2. Discussion of the decommissioning project.
- 3. Brief explanation of radioactivity and the biological effects of low 1cvel radiation exposure.
- 4. Emergency response actions.
Personnel will be required to score 80% or above on a written examination. Personnel will be required to complete annual ) requalification training. The requalification training will give special attention to changes in radiation protection, emergency planning and management policy. Personnel who require access to the Restricted Area to service and maintain equipment (e.g., vending machines, office equipment, etc.) or who have had unescorted access at other nucicar facilities within the last year, will be allowed to receive expedited project orientation training by:
- 1. Reading a copy of site specific information covering radiation protection, emergency response and access control procedure.
- 2. Signing a statement acknowledging their understanding of the information provided.
3.2.4.3 Non Radiation Worker Indoctrination All visitors who require access to radiologically controlled areas will receive indoctrination training. Visitor access to radiologically controlled areas will require approval of the Project Radiation Protection Manager or designee. This indoctrination training will-include:
- 1. The_ requirement that the visitor remain with the escort at all times and follow directions of the escort.
- 2. A description of the radiological nature and required
- controls of the area to be entered.
l
- 3. The purpose and proper use of dosimeters, including how to read self-reading dosimeters.
3.2 19
PROPOSED DECOMMISSIONING PLAN 6/28/91 SEC110N 3
- 4. Potential emergency situations and proper actions to take in such events.
3.2.4.4 Radiation Worker Training Personnel requiring unescorted access to radiologically controlled areas will be required to attend, as a minimum, Radiation Worker training. This training will provido workers with the knowledge and skills needed to work safely in radiologically controlled areas includingt radiation and high radiation areas, airborne radioactivity areas, radioactive material areas and contaminated areas. 1his training will be consistent with that outlined in NUREG 076), Appendix A, " Example Qualification Standard for Radiation Work Training" (Ref. 10). Radiation Worker training will include:
- 1. Biological effects of radiation and the risks associated with radiation exposure.
- 2. Information needed to comply with Radiation Protection procedures and respond properly to warnings and alarms under both normal and emergency conditions.
- 3. Information needed to ensure that individuals can maintain their own exposure ALARA and ensure that ALARA considerations are appropriately refiscted in decisions which affect the exposure of others,
- 4. Information needed to comply with Radiation Protection Program procedures.
- 5. Discussion of worker rights and responsibilities as identified ia 10 CFR 19.
- 6. Discussion of NRC Regulatory Guide 8.13 " Instructions concerning Pre natal Radiation Exposure" (Ref. 18).
- 7. Training in emergency response actions.
- 8. Discussion of radioactive and mixed waste minimization.
In addition to classroom training, each participant will be required to demonstrate their abilities in a practical factors session. This will include:
- 1. Properly don and remove a complete set of protective clothing (excluding respiratory protection equipment).
- 2. Read and interpret self reading dosimeters.
- 3. Read and interpret radiological survey maps,
- 4. Follow procedures to properly enter and exit a contaminated area, including use of proper frisking techniques.
- 5. Demonstrate understanding and compliance with a Radiation Work Permit.
3.2-20
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 Personnel will be required to score 80Y. or above on a written examination. Personnel who pass the written exam will be required to successfully complete the practical factors section. Personnel who fail exams will be evaluated to determine if additional training is needed, limited duty assignment is appropriate or disqualification is necessary. Radiation workers will be required to complete annual requalification training. The requalification training will give special attention to Radiation Protection Program changes, weaknesses observed in project personnel performance and lessons learned. Significant changes in Radiation Protection policies, requirements, techniques, procedures and equipment, as determined by the Project Radiation Protection Manager, will be disseminated in a timely ! manner to affected personnel or organizations through periodic l awareness presentations. l l Radiation Protection procedures will be developed to allow the l Project Radiation Protection Mantger, on a case by-case basis, to ; exempt personnel from Radiation Worker training. Exemptions will only be granted if the following conditions are met:
- 1. The individual is escorted by a qualified Radiation Worker or has documented proof of Radiation Worker training at another nuclear facility within the last year.
- 2. The exemption is valid only for the duration of the specific task for which access was approved.
- 3. The basis of the exemption and approval is documented.
3.2.4.S Specialized ALARA Trainina In addition to Radiation Worker training, separate and detailed instruction in advanced radiation work practices will be provided to those workers performing tasks that involve potential significant radiation exposure or quantities of radioactive material. This training will typically include workers involved in:
- 1. Operations which involve handling highly radioactive components that have the . potential for creating a significant airborne hazard.
! -2. Operations which require work in contamination containment devices.
- 3. Grinding, cutting or similar operations on highly radioactive systems, components or piping, j 4. Work activities that require the use of special tools and
( equipment for reducing exposures. l 3.2-21 l
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3
- 5. Special complex radiation work which involves skills and training beyond that covered in Radiation Worker training.
Specialized ALARA training may include mock ups, dry runs, pre job briefings and other special training classes. This training will normally be attended by all personnel involved with the task, including craf t supervision and Radiation Protection Technicians. The need for specialized ALARA training will be identified during ALARA reviews and/or Radiation Work Permit preparation. 3.2.4.6 Respiratory Protection Trainino Specialized respiratory protection training will be required for all personnel who use respiratory protection devices. Personnel using respirators in radiologically controlled areas will require Radiation Worker qualification. Respiratory protection training will be conducted in accordance with 10 CFR 20, NRC Regulatory Guide 8.15 (Ref.13) and NUREG-0041 (Ref. 14). Topics that will be addressed in respiratory protection training include, but are not limited to, the following:
- 1. Instructions that individuals are authorized to wear only the type of respirator for which they are fit tested and trained.
- 2. Discussion of the type of airborne contamination for which the respirators will provide protection.
- 3. Discussion of construction and limitations of respirator types.
- 4. Discussion of facial hair policy and use of approved eye wear.
- 5. Pre-use respirator inspection.
- 6. Instructions for proper donning and fit.
- 7. Emergency actions in the event of respirator failure including instructions to leave the area.
- 8. Practical demonstration of respirator inspection, donning and removal, in addition to classroom training, each participant will be required to demonstrate their ability in a practical factor session that includes:
- 1. Inspection of a respirator.
- 2. Donning a respirator and performing a negative pressure test.
- 3. Removing a respirator.
3.2-22
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 l Personnel will be required to score 80% or above on a written respiratory protection test. Personnel who pass the written exam and , are medically qualified will be required to successfully complete : the Practical factors and fit test session. Respirator users will be required to complete annual requalification training. 3.2.4.7 Radiation Protection Technician Training _ gad _ Qualification l l Radiation Protection personnel will be selected, trained and qualified to ensure that they have sufficient knowledge and practical abilities to implement the Radiation Protection Program effectively. Qualification criteria and job descriptions will be developed for all positions within the Radiation Protection organization. This qualification criteria will contain the elements outlined in Appendix 0 of Draft HUREG 0761 (Ref. 10) and will be l used to augment Radiation Protection Technician training, i All Radiation Protection Technicians will be required to participate in classroom and specific on-the-job training (0JT). The Radiation Protection Training Manual and implementing procedures will ensure that Radiation Protection personnel, who are selected, trained and qualified, have the knowledge and practical skills necessary to perform their work. Radiation Protection Technician qualification and training will include the following:
- 1. Radiation Protection Training procedures that specify Radiation Protection personnel qualification criteria, job
, descriptions, and responsibilities. l 2. Review and verification of resumes by the Project ! Radiation Protection Manager to ensure that personnel have sufficient education and/or experience in the job l functions which they will be assigned. Radiation Protection Technicians will be required to meet th education and experience levels specified in ANSI /ANS ' (Ref. 16).
- 3. Testing of Radiation Protection Technicians to verify appropriate knowledge level in radiation protection l theory, equipment, basic mathematics and -recognizing unusual situations involving radioactivity. The test will include representative topics listed in Appendix E of l
Draft NUREG 0761 (Ref. 10).
- 4. Training in Radiation Protection procedures, the operation ind limitations of survey and count room equipment and 3.2-23
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 methods to ensure proper record documentation and traceability.
- 5. Review of revisions to 10 CfR 20 and their impact on radiation protection activities.
- 6. Training in emergency response duties.
7 Review of major decommissioning work activities and potential radiological hazards that may be encountered. Radiation protection specialists (not qualified as ANSI 3.1 Radiation Protection Technician) performing unique radiation protection activities such as dosimetry, respiratory protection, bioassay, control point monitor, etc., will be provided specific task related training. This special'!cd classroom and/or on the-job training will be commensurate with ass gned duties and approved by the Project Radiation Protection Manager. Upon completion of required classroom training, Radiatic Protection personnel will complete on-thodob training in assigned duties. Successful completion of these duties will be documented by the responsible supervisor on the individual's qualification card. The responsible supervisor will ensure that training has been adequate by observation of on the-job performance. Annual Radiation Protection Technician refresher training, using a structured program approved by the Project Radiation Protection Manager, will be conducted. This training will be documented and may include a written examination. Additional training will be provided to Radiation Protection personnel if significant changes occur in Radiation Protection policy, requirements, techniques, procedures or equipment. This information will be disseminated to affected personnel or organizations through periodic awareness presentations and/or required reading. 3.2.4.8 M nsctive Waste Manacement Training Personnel assigned to the task of packaging, loading and shipping radioactive materials will be required to attend annual training commensurate with their assigned duties. This training will be in conformance with NRC IE Bulletin 7919, " Packaging of Low Level Radioactive Waste for Transport and Burial" (Ref. 20). The training will include instructions in all applicable Federal, State and local regulations and burial site requirements for classification, packaging, loading and shipping of radioactive materials. 1 3.2-24 l
-. - . ~ - _. . _ - .. . . - - - .._--__ - --- -
6/2M91 PROPOSED DECOMMISSIONING PLAN SECTION 3 3.2.4.9 flitt :.iion Prgig.tijan_$1nervisor Trainina and Qualification All project Radiation Protection supervisors will be trained and qualified in their positions. Radiation Protection training procedures will include supervisor job descriptions which specify qualification criteria and responsibilities. This qualification and training program will include the following: l 1. Review and verification of resumes by the Project Radiation Protection Manager to ensure that the supervisor has sufficient supervisory and technical experience in the area (s) for which they will be responsible. Supervisors designated as the alternate to the Project Radiation Protection Manager will meet the education and experience requirements established in NRC Regulatory Guide 1.8.
- 2. Training in Radiation Protection and decommissioning procedures associated with their area of responsibility.
3, Training in emergency response duties.
- 4. Periodic professional Radiation Protection training in the form of refresher courses, retraining, conferences or continuing education which enable Radiation Protection supervisors to keep abreast of current developments in the field.
1 3.2.4.10 Proiect Radiation Protection Manager Trainina and Oualification j The- qualification and training program for the Project Radiation Protection Manager will include:
- 1. Verification of prior edt. cation and experience as required by NRC Regulatory Guide 1.8.
- 2. Orientation on specific decommissioning plans, management organization and decommissioning project procedures.
- 3. Orientation on the specific design, systems and i radiological controls of the facility.
- 4. Training in emergency response duties.
5.- Periodic professional Radiation Protection training in the form of refresher courses, retraining, conferences or continuing education which enabic the Project Radiation- ' Protection Manager to keep abreast of current developments in the field. 3.2-25
. - - = - _ . - . . - - -
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 3.2.4.11 PSC Health Physicist Trainina and O nlification All PSC Health Physicists will be trained and qualified in their positions. Radiation Protection training procedures will include job descriptions which specify qualification criteria and responsibilities. This qualification and training program will include the following:
- 1. Review and verification of resumes by the PSC Radiation Protection Manager to ensure that the Health Physicists have sufficient technical experience in the area (s) for which they will be responsible. Health Physicists designated as the alternate to the PSC Radiation Protection Manager will meet the education and experience requirements established in NRC Regulatory Guide 1.8.
- 2. Training in Radiation Protection and decommissioning procedures associated with their area of responsibility.
- 3. Training in emergency response duties.
- 4. Periodic professional Radiation Protection training in the form of refresher courses, retraining, conferences or continuing education which enable Health Physicists to keep abreast of current developments in the field. ;
3.2.4.12 PSC Radiation Protection Manaaer Trainina and Oualification The qualification and training program for the PSC Radiation Protection Manager will include:
- 1. Verification of prior education and experience as required by NRC Regulatory Guide 1.8.
- 2. Orientation on specific decommissioning plans, management organization and decommissioning project procedures.
- 3. Training in emergency response duties.
- 4. Periodic professional Radiation Protection training in the form of refresher courses, retraining, conferences or continuing education which enable the PSC Radiation Protection Manager to keep abreast of current developments -
in the field. 3.2.4.13 Radiation Protection Tralnina Records The Radiation Protectioti Training Manual will specify the types of Radiation Protection training records to be maintained. Records will be maintained in accordance with regulatory requirements and company procedures. These training records will typically include: 3.2-26 i
e _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ i l 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3
- 1. Final written examination grade.
- 2. Final practical factors evaluation results.
- 3. Description of training completed satisfactorily, references to pertinent lesson plans, course outlines, syllabuses and other subject-specific descriptive information.
- 4. Documents indicating qualification verification (i.e.,
qualification cards). 3.2.5 Dose Control Radiation dose control is accomplished by controlling sources of radiation, controlling access to areas containing radioactive materials, measuring radiation exposures of workers, establishing exposure limits for workers and maintenance of an ALARA program. Specific elements of dose control include the following: ALARA Program Administrativo dose control Radiation Work Permits Area Definitions and Postings External dostmetry Internal dose control and monitoring Respiratory protection 3.2.5.1 ALARA Program All activities at fort St. Vrain involving radiation and radioactive material shall be conducted such that radiation dose to employees, contractors, and the general public are maintained ALARA. Project management will establish specific goals and objectives for the fort St. Vrain decommissioning project ALARA Program. The ALARA Program will be based on the guidance provided in Regulatory Guides 8.8 and 8.10 (References 11 and 12) to the degree applicable for decommissioning and dismantlement. 3.2.5.1.1 ALARA Prqaram Oraaniza'. ion and Responsibilities The PSC Radiation Protection Manager and the Project Radiation Protection Manager are cooperatively responsible to coordinate ALARA Program development and implementation. Specific responsibilities of the PSC Radiation Protection Manager and the Project Radiation Protection Manager, including their ALARA Program responsibilities, are listed in Section 3.2.2. 3.2-27
PROPOSED DECOMMISS10NilIG PLAli 6/28/91 SECTION 3 The actual implementation of specific ALARA actions, as incorporated into daily work activities, will be the responsibility of each individual manager, supervisor and worker. The ALARA Program will be supported by two levels of management providing oversight and direction. The working level of the ALARA Program organization will be the ALARA Committee and will be comprised of managers and representatives of various crafts at the supervisory level. Management oversight of the ALARA Program will be provided by the Decommissioning Safety Review Committee. The primary responsibilities of the ALARA Committee will be to review decommissioning work activities for effective dose reduction techniques and conformance with the Radiation Protection Program policies and procedures. Additional responsibilities will include reviewing the methods used for decommissioning, providing guidance and solutions for dose reduction and the approval of special 1:quipment and procedures used to reduce and maintain the overall project radiation dose ALARA. The ALARA Committee will be chaired by the PSC Radiation Protection Manager. The other members of the committee will be designated managers and/or supervisors involved in the decommissioning project. Members will have the appropriate authority and responsibility necessary to implement an effective ALARA Program. The ALARA Supervisor will implement the ALARA Program, serve on the ALARA Committee as a non-voting member and hold the position of Committee Secretary. ALARA Committee objectives will be as follows:
- 1. Ensure that ALARA policy, philosophy, commitments and regulatory requirements are integrated into all appropriate decommissioning work activities.
- 2. Establish overall ALARA Program goals for the decommissioning pro."-t.
- 3. Review and approve e ALARA budgets for specific project activities / tasks.
- 4. Review and evaluate individual and collective doses to determine the degree of success being achieved by the ALARA Program.
- 5. Initiate corrective actions, as necessary, to ensure accomplishment of ALARA Program objectives and goals.
- 6. Review cnd evaluate project activities / tasks that have dose estimates above specified action levels.
3.2-28
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3
- 7. Ensure that necessary resources are provided to achieve the goals and objectives of the ALARA Program.
- 8. Coordinate efforts of various functional groups (e.g.,
engineering, operations, technical support and Radiation Protection) to maintain radiation dose ALARA. The next higher level of management involvement in the ALARA Program will be the Decommissioning Safety Review Committee. This committee will review, evaluate and approve major decommissioning operations dealing with radioactive materials and radiological controls. This committee will also provide overall direction to the ALARA Committee for decisions involving financial solutions, administrative policy and decommissioning methods. 3.2.5.1.2 ALARA Trainina and Instruction Commitment to the principles of ALARA will be reflected in all radiation protection training. Training courses will be evaluated by the P,oject Radiation Protection Manager and approved by the. pSC Radiation Protection Manager to ensure that ALARA principles are incorporated into lesson plans. 3.2.5.1.3 [naineerina Controls l Engineering controls is the term used for the general class of , devices and associated methods used to reduce the exposure of personnel to radiation and radioactive material. Engineering controls typically include temporary shielding, engineering access controls, process instrumentation, control of airborne radiation sources, remote surveillance equipment, control of surface contamination, and other work improvement techniques.
" Temporary Shielding" . will be evaluated during the planning phase for activities involving high dose rates, such as core component df sinantlement and removal. The use of temporary shielding will -'
continue to be evaluated during the implementation - phase. The decision of whether to use temporary shielding will be based on considerations such as:
- 1. The effectiveness of providing shielding for'the component (radiation source) or shielding between the source and the worker (shadowshields).
- 2. The effectiveness of providing partial shields such as for
" radiation streaming", or "high level" sections of piping, drains, sumps, etc.
- 3. Estimated cose savings by the use of shielding.
3.2-29
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3
- 4. Estimated dose expended during shleiding installation.
- 5. Projected cost of installing and removing shielding, including the cost due to delay of project, if applicable.
" Engineered Access Control" will be evaluated and used to limit access to liigh Radiation Areas where it is not practical to provide continuous positive control (e.g., Radiation Protection Technician stationed at the ingress / egress point). Examples of engineered access controls include:
- 1. Inaccessible barriers
- 2. Locked gates or doors
- 3. Barriers with flashing light (s)
- 4. A combination of the above
" Process Instrumentation" for systems (e.g., the PCRV water filtration system, radwaste processing systems, etc.), will be reviewed by the Project Radiation Protection Staff for instrument location and layout, including such concerns as:
- 1. General accessiblitty and associated radiation exposure.
- 2. Potential radiation exposure due to operation of the system.
- 3. Potential- radiation exposure due to servicing and maintaining the system / process.
" Control of Airborne Radiation Sources" will be considered for work activities _ that hava this potential for producing airborne radioactivity (e.g., cutting and grinding operatinns). Engineering controls to confine and/or control the source will be evaluated.
Examples of engineering controls for airborne radiation sources include:
- - 1. Existing plant ventilation / filtering systems.
- 2. Auxiliary ventilation / filtering systems for contaminated components and for machining and grinding.
- 3. Contamination control containments.
- 4. Purification systems.
- 5. Decontamination equipment.
- 6. Wet handling of highly contaminated equipment, such as
- PCRV components.
l 7. Air sampling / monitoring instruments located to provide a L quick indication of elevated airborne levels.
" Remote Surveillance Equipment" (e.g., TV monitors, audio equipment, "as-installed" photographs and radiation monitors) will be evaluated 3.2-30
6/28/91 PROPOSED DLCOMMISS10NING PLAN SECTION 3 for use during decommissioning activities that have the potential for producing high radiation or airborne radioactivity areas. Such equipment, when used, will allow personnel to evaluate the radiation or airborne lovels, the layout of the area and activities in the area without being exposed to the radiological conditions of the affected area.
" Control of Surface Contamination" will be evaluated and used to control and contain the spread of contamination and prevent the spread of radioactive materials to uncontrolled areas. Examples of methods used to control the spread of contamination include:
Containments Glove bags Surface decontamination Drip pans Air curtains and plastic coverings Leak control Strippable coatings Other engineering controls that may promote work efficiency and reduce radiation dose to workers will be evaluated. Examples of other engineering controls include: Adequate lighting Adequate ventilation Adequate working space Ease and quickness for installing / dismantling temporary equipment, such as scaffolding & insulation Heans of accessibility, such as working platforms, cat walks, and fixed ladders Removal of components to remote areas, whcre shielding and special tools are available in low radiation areas 3.2.5.1.4 ALARA Program Goalt Each major task / activity during decommissioning will be assigned an ALARA (person Rem) goal. Each goal will be established based on the anticipated person-hours, dose rates and dose saving methods employed. The goals will not be an estimate of the dose to be expended, considering the dose rates and person-hours, but will be a goal that requires planning and proper execution of dose-saving methods on the part of all personnel involved in the task. 3.2-31
- , ,-- - - - - . ~ - - - - _ - - _ - . . - - _ - . - _ . . -
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 The estimated dose from major tasks will be combined with minor and/or routine tasks (if necessary) to determine the annual and 4 total dose goals for the Decommissioning Project. Major. activities / tasks 'wil'1 be evaluated during the planning phase. This will involve input from design engineers, the Project Radiation Protection Staff and implementing supervisors to determine an estimated ALARA goal for the task. ALARA goals will be adjusted as additional job reviews, ALARA reviews and task modifications are
-performed.
l Prior to the start of specific activities / tasks, ALARA goals t ill be ] reviewed and approved by the ALARA Committee and the Decommiss,,ning ) Safety Review Committee, as appropriate. 3.2.5.1.5 ALARA Job Reviews ALARA ' job -reviews will be conducted for each activity / task as_ an integral part of Radiation Work Permit preparation. (See Section 3.2.5.3) The scope - and _ detail of the ALARA -job review, and the individuals designated for. review will depend on the complexity of the. task, the expected- radiological conditions and the estimated doses.
~ Cost benefit will be considered in ALARA job reviews where significant costs may be incurred to- reduce - the estimated dose.
-_These costs may be incurred by such factors as temporary ' shielding, remote tools. and/or : surveillance, ventilation- controls, decontamination, etc.
Pre-job briefings 'and ~ mock-up traini.ig -will be conducted for
- complicated- or high; dose jobs, as required by implementing procedures and/or the specific Radiation Work Permit.
In-progress. ALARA reviews will be required if radiological conditions change significantly , are unexpected or if the projected 4ccumulated idose differs significantly -from the estimated value _(e;g.,. greater than twenty-five percent). Post-job ALARA ' reviews will be conducted. at the completion of activities / tasks lfor which significant ALARA- planning measures were required or accumulated person-Rem exceeded specified action levels, g Post-job reviews will encourage all personnel- involved in the project activities to provide _ input regarding the effectiveness of- , methods.used to perform the work. This input will be evaluated for 3.2-32
l 6/28/91 PROPOSED DEC0HMISS10NING PLAN SECTION 3 future work activities to improve work conditions and maintain worker dose ALARA. 3.2.5.1.6 ALARA Work Practicci in order to complete each individual task and minimize personnel exposures, the following work practices, as a minimum, will be implemented:
- 1. Pre-job briefings will be held with craft and Radiation Protection personnel to assure that ALARA practices have been adequately factored into the work packages for completing tasks.
- 2. Personnel exposures will be monitored on a regular basis for potentially high exposure tasks to identify any irregularities that may indicate excessive personnel exposures. In the event that an irregularity is found it will be investigated immediately and corrective actions implemented.
- 3. Tag lines will be attached and used when rigging and lifting high exposure rate components (e.g., steam generators) from the PCRV to kets workers as far from the source as possible.
- 4. Only essential personnel will be allowed inside the Reactor Building when high exposure rate components are being removed. Casual observers will not be permitted.
- 5. Bagging techniques will be designed for quick installation.
- 6. Long-handled wipe tools will be used when appropriate to wipe down wet components removed from the PCRV.
- 7. Shadow shields (lead blanket curtains or equivalent) will be used, where appropriate, to reduce radiation fields at the work stations. .
The radiation levels from activated structures inside the PCRV will be measured as segments of the reactor are dismantled. The measurement of exposure rates at the individual work stations will be performed and compared with calculation prior to commencing each individual task. Adjustments will be made in exposure projections as necessary. Temporary shielding will be used at the work stations 1 3.2-33
L PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 to minimize personnel exposures based on actual exposure rate measurements. 3.2.5.1.7 Administrative Controls During the Project planning phase, specific work packages and the project as a whole, will be reviewed by the Radiation Protection staff to ensure that adequate administrative controls and Radiation Protection hold points are included. The scope and detail of the controls and hold points will be a function of the estimated radiation dose rate levels and the complexity and duration of the work activities. Work activities that have the potentisl for high exposure rates (e.g., greater than 500 mrem /hr) or high estimated dose for the task (e.g., greater than 10 person-Rem) will be reviewed and approved by the ALARA Committee. Work activities, that if not performed strictly in accordance with administrative controls could potentially produce personnel exposure in excess of regulatory limits, will require review and approval by the PSC Program Manager for Decommissioning and/or the Decommissioning Safety Review Committee. 3.2.5.1.8 ALARA Suaaestion Proaram An ALARA suggestion program will be developed to ensure that all personnel have the opportunity to participate in identifying potential ALARA concerns or recommendations to support the project's dose reduction efforts. All suggestions will receive review and response by the Radiation Protection staff to maintain open communications on ALARA issues. ALARA Suggestion Program implementing procedures will address: submitting ALARA suggestions; reviewing, evaluating and approving ALARA suggestions; and implementing and tracking ALARA suggestions. 3.2.5.1.9 ALARA Proaram Evaluation and Aooraisal ALARA Program effectiveness will be monitored and evaluated on a continuing basis to determine appropriateness and effectiveness. A variety of feedback mechanisms will be in effect to provide information for these evaluations. These mechanisms will include pre-job, on-going and post-job reviews, and trending of ALARA Performance Indicators. 3.2-34
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 ALARA Performance Indicators will be used to monitor and trend a variety of indicators to identify those areas where the ALARA program is performing effectively and also where problems may be occurring in the Radiation Protection Program. The need for additional ALARA Performance Indicators will be identified through the review of . ALARA suggestions, radiological occurrences, radiological incidents, etc. Typical ALARA Performance Indicators include, but are not limited to: Respirator usage as compared to the number of entries to controlled areas Personnel contaminations Collective and individual doses Number of positive bioassay results 3.2.5.2 Administrative Dose Control Administrative radiation dose controls will be implemented to ensure personnel do not exceed regulatory dose limits, to ensure an equitable distribution of dose among project workers with similar jobs and to ensure that the collective dose to workers i: ALARA. The Radiation Protection Manual implementing procedures will detail administrative dose control requirements and activities. These ! procedures will include, but not be limited to, the following elements:
- 1. An approval system by various levels of supervision and management.which controls both planned and actual doses to individuals as they progressively (incrementally) . approach regulatory or established administrative limits.
- 2. Permission to exceed the lower administrative limits (e.g., 500 mrem / quarter and/or 1000 mrem / year whole body dose for adults) will require approval of the-individual's l supervision and the Project Radiation' Protection Manager.
- 3. Permission to exceed the higher administrative limits (e.g., 2000 mrem / year whole body dose for adults) will.
require approval of the individual's supervision, the Project Radiation Protection Manager, the PSC Radiation l Protection Manager, and the PSC Program Manager for l Decommissioning. l 4. The Project Radiation Protection Manager's approval to i exceed an administrative limit will be based on a l determination that the dose to be received by the individual is ALARA. 3.2-35
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3
- 5. Administrative controls will be in place and all personnel will be instructed and trained in these controls to ensure that any activity can be and will be stopped, if necessary, to re-evaluate the evolution and ensure no excessive doses are incurred.
- 6. Specific administrative limits and guidelines for exposure to the unborn, visitors, minors, etc.
- 7. Guidelines 5 policies governing emergency dose authorization and methods for handling overexposures.
3.2.5.3 Estdiation Work Permits Radiation Work Permitt, (RWPs) will be used for the administrative control of personnel entering or working in areas that have, or potentially have, radiological hazards present. RWPs will summarize the Radiation Protection controls established as part of job planning and will be detailed enough to deal with changing (or - potentially changing) radiological conditions expected during the course of the work. RWPs will use current radiological survey information when establishing dose control measures and will specify any special survey requirements prior to, during and after the work activity. An RWP will be required for the following:
- 1. Entry into or work in a radiologically controlled area.
- 2. Entry into Radiation, High Radiation, or Radioactive Materials Areas located outside a radiologically controlled area.
- 3. Activities involving equipment, controls, or instrumentation containing or. suspected of .containing radioactive material which are located outside a radiologically controlled area.
- 4. When determined _ by Radiation Protection that radiological controls in the form of an RWP are appropriate.
The RWP process will provide a systematic method to evaluate i radiological conditions under which decommissioning work activities will be accomplished, specify radiation protection requirements and' ensure that required worker briefings are given. Acceptable radiation work practices will be described and sufficient Radiation ' Protection Technician coverage assigned to assure worker protection and ensure that worker dose is maintained ALARA. RWPs will also provide a method to record doses for each individual by major job or task. The recorded RWP doses will also allow dose trend analysis i and will frequently provide workers with their current dose status. 3.2-36
i 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 The specific methcds used for dose accountability and trending will be prescribed in the Radiation Protection Manual implementing procedures. Appropriate management approvals for RWPs involving significant projected total dose will be established in conjunction with the ALARA Program. RWP preparation and approval will be specified in Radiation Protection implementing procedures. RWPs will typically provide the following information:
- 1. Description of job or activity to be performed.
- 2. Anticipated radiological conditions i -luding, as applicable, contamination levels, radiation level s ,
airborne radioactivity levels.
- 3. Reference to, or a copy of, dose rate and contamination level survey maps.
- 4. Number and identification of personnel assigned to the job or activity (if appropriate).
- 5. Monitoring requirements during the job, such as constant radiation protection coverage, intermittent coverage, air monitoring, etc..
- 6. Special instructions and equipment to minimize exposure to radiation and contamination.
- 7. Protective clothing and equipment requirements.
- 8. Personnel dosimetry requirements (e.g., whole body, extremity).
- 9. Estimated exposure time and dose (person-Rem) to complete the task.
- 10. Actual exposure time, dose and other information obtained durlig the task.
- 11. ALARA pre-job briefing elements.
RWPs will normally be initiated by the group responsible for the job to be performed. The RWP request will typically include the following information:
- 1. Job description, work package number and if appropriate the purpose of the task.
- 2. Location (s) of work.
- 3. Estimated time to complete the job including, if appropriate, crew size and work location.
- 4. Exposure estimate.
RWPs will be documented in a legible and easy to comprehend format, and will be readily accessible for workers' review. All perscnnel assigned to a RWP will be required to read and sign the RWP - verifying they will comply with its requirements. It will be 3.2-37 I
PROPOSED DECOMMISSICNING PLAN 6/28/91 SECTION 3 expected that radiation workers be in strict compliance with RWP requirements. Willful or habitual disregard of RWP instructions will be cause for disciplinary action. RWPs will be classified as Standing or Special, as determined by Radiation Protection Supervision. Standing RWPs will be used for the performance of routine activities such as observation, inspection, operator rounds or laundry operations where radiological conditions are stable. The Project Radiation Protection Manager will approve all Standing RWPs. Job specific RWPs will be used for the performance of a defined activity in specific locations. Radiation Protection Manual implementing procedures will specify how RWPs will be maintained and retained. These records will typically include copies of RWPs, RWP sign-in logs and ALARA review documentation. 3.2.5.4 Area Definitions and Postinas Radiological postings will be provided at the entrance and boundaries of radiological areas to advise workers of radiological hazards. Methods will also be provided to clearly distinguish radioactively- contaminated systems and indicate any special precautions required for work on such systems. Informational postings may also be used to provide additional radiological instructions to workers. It is the responsibility of each worker to observe the radiological postings and comply with the indicated requirements. 3.2.5.4.1 Radioloaical-Postinas A Radiation Area is any area, accessible to personnel, in which there exists radiation, originating in whole or in part within licensed material, at such levels that a major portion of the body could receive in any one-hour a dose in excess of 5 mrem, or in any 5 consecutive days, a dose in excess of 100 mrem. Radiation Areas will be posted " CAUTION -RADIATION AREA". Radiation area boundaries will= be designated by the use of barriers, walls, ropes, markings and/or signs. A Hiah Radiation Area, as defined in the Decommissioning Technical Specification 5.8.1, is an area where whole body dose rates exceed 100 mrem /hr at 45 cm (18 inches) from the radiation source. These areas will be posted " CAUTION - HIGH RADIATION AREA". High Radiation Areas will be barricaded and conspicuously posted and entrances will be controlled by requiring a Radiation Work Permit. 3.2-38
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 Personnel permitted to enter High Radiation Areas will be provided with or accompanied by one or more or the following:
- 1. A survey instrument which continuously indicates the area dose rate.
- 2. A radiation monitoring instrument which continuously integrates the radiation dose rate in the area and alarms when a preset cumulative dose is received (e.g., Digital Alarming Dosimeters (DADS)). Entry is permitted only after the area dose rate has been made known to personnel.
- 3. A Radiation Protection Technician equipped with a survey meter. The Radiation Protection Technician will be responsible for performing radiological monitoring at the frequency specified on the Radiation Work Permit and for providing positive control over the activities within the area.
High Radiation Areas, as specified in Decommissioning Technical Specification 5.8.2, in which the dose rate is greater than 1000 mrem /hr at 45 cm (18 inches) from the radiation source will be provided with locked enclosures to prevent unauthorized entry into such areas. Keys to such areas will be under the administrative control of Radiation Protection supervision. High radiation enclosures will remain locked except during periods of access by personnel under an approved RWP which will specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in the area. In lieu of the stay time specification of the RWP, continuous Radiation Protection Technician surveillance (either direct observation or remote by use of closed circuit TV cameras) will be used to provide positive control over the activities within the area. Certain areas accessible to personnel with radiation levels of greater than 1000 mrem /hr may be located within larger areas where ~~ m enclosure exists or can be reasonably constructed for purposes of locking individual areas. In this case, the area will Le roped off, conspicuously posted, and a flashing light will be activated as a warning device. A Contamination Area is an area accessible to personnel with loose surface beta-gamma radioactivity in excess of 1000 dpm/100cm2 or loose surface alpha radioactivity exceeds 20 dpm/100cm2, Contamination Areas will be conspicuously ported with the radiation symbol and the words "CAVTION - CONTAMINATION AREA". Contamination Area boundaries will be designated by use of barriers, walls, ropes, markings, signs and/or step off pads. 3.2-39
PROPOSED' DECOMMISSIONING Pl.AN' 6/28/91 SECTION 3-An A'irborne -Radioactivity Area is an area where airborne r radioactivity is present, at concentrations greater than the Derived i Air Concentration _ (DAC) as specified in 10 CFR 20, or 12 DAC-hours over 40 hours. Airborne Radioactivity Areas will be conspicuously posted at the entrance with " CAUTION - AIRBORNE RADI0 ACTIVITY AREA". Airborne Radioactivity Area boundaries will be designated by use of barriers, walls, . ropes, markings and/or signs. l A Radioactive Materials Area is an area or room in.which radioactive material is used or stored in -an amount exceeding 10 times (100 times for Uranium _ and Thorium) the quantity specified in 10 CFR 20. Radioactive Material Areas will be posted " CAUTION -RADI0 ACTIVE MATERIALS AREA". Radioactive Materials Area boundaries will .be ;
- designated by the- use of barriers, walls, ropes, markings and/or i L signs.
- A Radioloolcally Controlled Area (RCA) is defined as any of the above defined areas, including a Radiation Area, High Radiation Area, : Contamination Area, Airborne Radioactivity Area, or l Radioactive Materials Area. RCA boundaries will be defined by the t
use of barriers, walls, walls, ropes, and markings. The boundaries will be clearly marked with posted signs that identify the type (s) i of- RCA within the boundaries, and entrance and exit points will normally be. posted. 3.2.5.4.2 Informational Postinas A Hot Particle Area will be established to identify areas where - di.screte - particles with high specific activity are. located. Hot particle areas will - be' contained and posted within a contaminated area as " HOT. PARTICLE ~ AREA". A . ' Hot Soot is where localized dose rates near an item are much L greater (e.g., fivef times) than the -general area whole body dose rates. These locations are posted " CAUTION - HOT SPOT". . - A7 Restricted Area is. any area to- which access is controlled for-purposes of protection of' individuals from exposure to. radiation and L- radioactive material. An Unrestricted Area is any area to which _ access' is not controlled for purposes of protection of individuals from exposure to radiation and radioactive materials. Except as authorized per 10 CFR ' 20,
-radiation ' levels in " Unrestricted Areas" will not exceed levels that, if an individual were-continuously present, would result in an 3.2-40 . . _ , . . . _ . , _ . . . . .J_.. ., m . _ , , .. , . . . _ . . _ _ _ _ _ . _ _ . _ _ _ . . _ _ _ _ . . . . _ _ _ _ . . , _ _ . . .
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 individual receiving in excess of 2 mrem in any one hour or 100 mrem in any seven consecutive days. A low Dose Waitina Area (LDWA) is established to identify areas where dose rates are lower than other locations within the work area. When practical, workers should be directed to remain in a LDWA unless they are actually needed at the work location. LDWAs will be posted; " LOW DOSE WAITING AREA". 3.2.5.5 External Dosimetry 3.2.5.5.1 General Considerations l-External radiation dose monitoring will be accomplished through the use of thermoluminescent dosimeters (TLDs) and sel f-reading dosimeters (SRDs). The official record of external dose to beta, gamma and neutron radiations will normally be obtained from TLDs. SRDs will be used as a means for tracking dose between TLD processing periods and may also be used as a back-up to the TLD. TLDs will be processed at a frequency to ensure personnel dose limits are not exceeded. A contract dosimetry service will supply TLDs during the decommissioning project. The dosimetry laboratory will be accredited -by the National Voluntary Laboratory Accreditation Program (NAVLAP) for the radiation type and energy expected to be monitored. Self-reading dosimeters will meet the requirements specified in NRC Regulatory Guide 8.4, " Direct-Reading and Indirect-Reading Pocket Dosimeters", (Ref. 21) February 1973. l 3.2.5.5.2 Monitorina Whole Body Dose All project workers will be required to wear external radiation l monitoring devices whenever they enter radiologically controlled areas. SRDs will be read prior to their use and periodically tFeceafter by the wearer. If a SRD is off-scale or lost under l cond;tions such that a high dose was possible, the individual's TLD l will be promptly processed and the individual will be denied access I to radiologically controlled areas. The TLD and SRD will normally be worn on the trunk of the body between the neck and waist in close proximity to each other. Under certain conditions, where the chest or trunk may not be the location of highest whole body dose, dosimetry -devices may be relocated. Radiation Protection Manual implementing procedures will specify criteria for relocating whole body dosimetry. 3.2-41
i PROPOSED DECOMMISSIONING PLAN 6/28/91 ; SECTION 3 ! The use of multiple whole body dosimetry will be evaluated whenever work is to be performed in a non-uniform radiation field and that portion of the body which will receive the highest dose is not easily determined, in these cases, multiple sets of dosimeters will be worn on those parts of the body expected to receive the highest dose. Guidance for conducting the evaluation and criteria for determining when multiple dosimetry is required will be provided in Radiation Protection Manual implementing procedures. RWPs will be used to communicate dosimetry requirements to the workers. 3.2.5.5.3 Monitorina Extremity Dose j Extremity monitoring devices will be used whenever is likely to receive a dose in excess of 25% of the quarterly extremity dose limit of 18.75 Rem. Guidance for evaluating the need for extremity monitoring will be provided in Radiation Protection Manual implementing procedures. 3.2.5.5.4 MonitorinqJLkin Oose Monitoring of the skin of the whole body will normally be accomplished utilizing the whole body TLD. Calculation of skin dose due to contamination or " hot particles" will be performed in . accordance with acceptable models and equations identified in NRC Bulletins, Regulatory Guides and in published technical literature. The methods for calculating and documenting skin dose due to contamination and " hot particles" will De provided in Radiation Protection Manual implementing procedures. l 3.2.5.5.5 Dosimetry Ouality Control
- Periodic quality assurance checks of vendor supplied dosimetry will be conducted. Analysis of the results will be performed as i described- in ANSI N13.ll, " Criteria for Testing Personnel Dosimetry", (Ref. 23) 1983. These checks will be addressed in Radiation Protection Manual implementing procedures._ in addition, SRD results will be compared to TLD results. Each discrepancy greater than 25% for doses over 100 mrem will be evaluated. The evaluation will include consideration of factors such as energy dependence of the device used, survey results, exposure times, doses of other personnel performing similar work, location of devices worn on the body and clerical errors.
3.2-42
l 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 3.2.5.6 Internal 0qtisetry Contrq.1 and lionitorina 3.2.5.6.1 General Considerations Internal radiation dose is inherently mere difficult to measure than external radiation dose, but is generally much easier to prevent. Therefore, the major emphasis will be placed on preventing internal radiation dose, provided it is consistent with the goal of keeping total effective dose ALARA. The primary methods for controlling intake of radioactive material into the body will be identifying and minimizing sources of airborne radioactivity and applying engineering controls to reduce airborne radioactivity concentrations. The use of respiratory protection will serve as a secondary method of control. Administrative controls and limits will be established to minimize intakes of radioactive materials. Radiation Protection Manual implementing procedures will be developed to conduct a routine bioassay program including criteria for the performance of bioassay, dose tracking and methods for data analysis and interpretation. The bioassay program will be based on NRC Regulatory Guide 8.26, " Application of Bioassay for Fission and Activation Products", (Ref. 24) September 1980, NRC Regulatory Guide 8.9, " Acceptable Concepts, Models, Equations and Assumptions for a Bioassay Program", (Ref, 25) September 1973 and NR. (egulatory Guide 8.32 " Criteria for Establishing a Tritium Bioass . Program", (Ref.
- 26) July 1988.
3.2.5.6.2 Whole Body Counting Whole body counting will be the primary method used to determine the identity and quantity of gamma emitting isotopes in the body at any given time. Radiation workers will receive, as a minimum, a baseline and annual whole body count, in addition, personnel will receive a whole body count after a suspected intake of radioactive materials. Radiation Protection Manual implementing procedures will provide guidance on whole body counter operation, calibration and quality control. 3.2.5.6.3 Indirect Bioassay Indirect bloassay (in-vitro) measurements will be made, as necessary, to monitor for alpha and beta emitting radioisotopes. This method of bioassay will typically be used only for isotopes which cannot be determined by whole body counting or when additional 3.2-43
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 information on an intake is required. Urinalysis will be used to assest personnel intakes of tritium. Radiation Protection Manual implementing procedures will- include criteria for indirect bloassay and methods for data analysis and interpretation. 3.2.5.7 Respiratory Protection Proaram The Respiratory Protection Program will be established in accordance with 10 CFR 20 and consistent with the guidance of NUREG-0041 (Ref. 14). The primary objectives of the Respiratory Protection Program are personnel safety and limiting the inhalation of airborne radioactive materials. Engineering controls will be applied to minimize concentrations of radioactive materials whenever l practicable. When engineering controls are not practicable, other , controls such as increased surveillance, limitations of working l times or use of respiratory protection equipment may be appropriate. l The Respiratory Protection Program will include the following l elements: l
- 1. A written policy statement and standard operating procedures.
- 2. Guidance on proper selection of equipment, based on the
! hazard.
- 3. Proper training and instruction to users.
l 4. Proper fitting, use, cleaning, . storage, inspection,
- quality assurance and maintenance of equipment.
- 5. Appropriate surveillance of work conditions.
! 6. Regular inspection and evaluation to determine continued l" program effectiveness.
- 7. Program responsibility vested in one qualified individual.
- 8. An adequate medical surveillance program for respirator L users.
! 9. Use of only Bureau of Mines /NIOSH - certified or NRC l authorized equipment. j 10. Maintenance of a bioassay program. l ) 3.2.5.7.1 Proaram Administration l-The Respiratory Protection Program will be administered by a Radiation Protection staff member designated by the Project Radiation Protection Manager. Qualifications of this individual will satisfy the requirements of ANS/ ANSI 3.1. Program administrative responsibilities will typically include the following: 3.2-44 i
l 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3
- 1. Provide overall program development, technical direction and the evaluation of program effectiveness.
- 2. Provide technical guidance for the control of airborne radiological contaminants.
- 3. Develop procedures, training materials and directives related to the program.
- 4. Conduct routine overviews of the program for compliance with policy, procedures and regulations.
3.2.5.7.2 Egipir1 Lor User Qualification Respirator user qualification criteria will include:
- 1. Annual examination by a physician to establish physical and psychological capabilities necessary to perform tasks using a respirator. A medical re-evaluation will be performed annually.
- 2. Successful completion of respiratory protection training as described in Section 3.2.4, " Radiation Protection Training and Qualification."
- 3. Successful quantitative fit test prior to the use of respirators requiring a facepiece-to-face seal on an annual basis.
- 4. No facial hair between the face and the sealing surface of the respirator and no facial hair interfering with valve function of the respirator.
3.2.5.7.3 Bioassay Bioassay techniques will be used to determine the amount and type of radionuclides in the body as an evaluation of the effectiveness of the Respiratory Protection Program. 3.2.5.7.4 Respiratory Protection Eouipment Description and Selection The selection of respiratory protection equipment will be based on work area survey data and/or expected airborne contamination levels. The need for respiratory protection will normally be determined and prescribed by Radiation Work Permits. All work tasks in contaminated areas will be evaluated for respiratory protection requirements. Special attention will be given to the requirement for respiratory protection when the work activity / task involves any of the following operations: Thermal cutting Concrete scabbling 3.2-45
PROPOSED DECOMMISSIONING PLAN 6/28/91
-SECTION 3 Welding Grinding Concrete demolition Respiratory protection equipment will be selected to provide a protection factor greater than that required for the expected peak concentration of airborne radioactive materials in the work area.
Assigned protection factors will not exceed those specified in 10 CFR 20. If the selection of a respiratory protection device is inconsistent with the goal of keeping total effective dose AtARA, consideration will be given to alternative controls or respiratory protection equipment with a lower protection factor. 3.2.5.7.5 Sunolied Air Respiratory Eouicment Breathing air may be supplied to respirators from compressed air cylinders,- air compressors or the plant breathing air system. All sources of compressed breathing air will meet the requirements for Grade 0 breathing air as specified in ANSI /CGA G-7.1, " Commodity Specification for Air", (Ref. 27) 1989. 3.2.5.7.6 Eouloment_ Inspection and Maintenance Requirements and techniques for inspection and maintenance of respiratory protection equipment will be contained in Radiation Protection Manual implementing procedures. Inspection and maintenance will- be performed in accordance with manufacturers' and regulatory. requirements. ! Respirators will be maintained and issued in a N10 Sit certified ! configuration'. The- certification for a respirator will be automatically voided if the respirator is not the same in all respects as certified by NIOSH or if the respirator is not maintained in a certified condition. Acceptable methods of cleaning . respiratory equipment will be performed in accordance. with manufacturers' specifications and Radiation Protection Manual implementing procedures. l 3.2.5.7.7 pjuality Assurance (0A) L E Periodically, respirators will be randomly selected to verify they l- have been - properly cleaned and inspected. Respirator users will j also' be randoinly selected for whole body counts to verify program effectiveness. in addition, if there is an indication of equipment 3.2-46 l
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 failure or improper use (e.g., positive nasal smear), the respirator user will be whole body counted. 3.2.5.7.8 DAC-Hour Trackina Administrative and engineering controls will limit the intake of radioactive materials to levels that are ALARA. An administrative limit of 2 DAC-hours in any one day and 10 DAC-hours in any one week will be established. Radiation Protection Manual implementing procedures will provide guidance on DAC-hour determination and tracking. 3.2.6 Radioactive Material Controls Radioactive material controls will be established to provide for positive control of radioactive material, prevent inadvertent release of radioactive materials to uncontrolled areas, ensure personnel are not unknowingly exposed to radiation from lost or misplaced radioactive material and minimize the amount of radioactive waste. material generated during the decommissioning. Radioactive material is defined as material activated or contaminated by the operation or decommissioning of Fort St. Vrain and byproduct material procured and used to support the operation or decommissioning of fort St. Vrain (e.g., calibration sources, check sources and radiography sourcesJ.
-The Radwaste Supervisor and the Radiation Protection Operations Supervisor will share the responsibility for the radioactive material controls.
Detailed radioactive material controls will be described and implemented by the Padioactive Waste-and Radiation Protection Manual implementing procedures. Specific radioactive material controls include the following: l l Receipt of radioactive material Identification of radioactive material Control and movement of radioactive material Storage of radioactive material Accountability and inventory of radioactive sources Release of materials for unrestricted use Control of materials entering radiologically controlled areas Preparation of radioactive materials for shipment Radioactive liquid and gaseous release 3.2-47
t PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 3.2.6.1 Receipt of Radiqactive Material Implementing procedures for the purchase and receipt of radioactive material will include, but not be limited to the following requirements:
- 1. Requests for radioactive materials will be submitted to the Project Radiation Protection Manager for license verification and inventory check.
- 2. Personnel who initiate purchase orders for radioactive materials will be required to specify that shipments be marked " Attention: Project Radiation Protection Manager".
- 3. Picking up, receiving and opening packages identified as containing radioactive material will be performed in accordance with 10 CFR 20. The Project Radiation Protection Manager, or designee, will be notified as soon as possible after receipt of any package. The external surface of the package will be inspected and surveyed.
Shipping damages and other discrepancies will be documented and Radiation Protection Supervision notified. 3.2.6.2 Identification of Radioactive Material Implementing procedures will be developed to specifically address the identification of radioactive material. Each container in which radioactive material is transported, stored, or used in quantities exceeding those given in 10 CFR 20, will bear a durable, clearly visible label, identifying the radioactive contents. This label will contain, as a minimum, the radiation caution symbol and the words; " CAUTION OR DANGER - RADI0 ACTIVE MATERIALS". The label will provide sufficient information to permit individuals handling or using the material, or working nearby, to take precautions to avoid or limit their exposure. Requirements for radioactive material labeling may be excepted if:
- 1. The material is uniquely identified for use as radiological protection equipment (e.g., respirators, protective clothing, etc.).
- 2. The material is under the direct control of personnel trained as radiation workers who are aware of the contents and the associated radiological hazards.
- 3. The material consists of radiological samples or sampling equipment in the custody of Radiation Protection personnel.
- 4. The material is packaged and labeled in accordance with DOT regulations while awaiting transport.
I 3.2-48
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3
- 5. The material is contained in permanently -. installed equipment and/or potentially contaminated systems.
Radiation level posting requirements shall remain applicable. Radioactive material storage areas will be posted as " CAUTION - RADI0 ACTIVE MATERIAL AREA", in addition to postings required for l radiation and contamination, as indicated in Section 3.2.5.4. 3.2.6.3 Control and Movement of Radioactive Material in the Rgstricted Area Implementing procedures will be developed to specifically address the. control and movement of radioactive materials within the l Restricted Area. Radioactive material removed from contaminated ; areas will be contained, surveyed and labeled to allow appropriate ' control of radioactive material. Radioactive liquid samples or sources- will be properly contained and will be transported by, or escorted by, Radiation Protection qualified personnel. Materials which are to be prepared for shipment or storage will be packaged in containers suitable for shipping or storage, as applicable. The materials and packages will be surveyed for radiation and contamination - levels and the package appropriately L labeled to reflect those levels. External surfaces of containers or l wrappings containing radioactive material will_ be surveyed to ensure that loose surface contamination levels meet the unconditional release criteria, unless specifically exempted by Radiation Protection Supervision. Whenever material or equipment is transferred from one location to another location within a radiologically controlled area, it will meet the radiation and contamination limits of each area through l which it passes or be under the control of Radiation Protection personnel. t . ! Procedures for the control and movement of radioactive material will l include, .but not be -limited to, the following provisions:
- 1. Guidelines for monitoring and handling radioactive material (e.g., quantities, geometries and estimated concentrations will be used' to estimate radiation levels and container requirements).
- 2. Unique features will be used (e.g., yellow plastic bags, yellow and magenta tags, etc.) to clearly identify the
! physical and radiological parameters of the material. 3.2-49
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3
- 6. Radioactive material will not be stored outside of the Restricted Area unless specifically approved by the Project Radiation Protection Manager.
Radioactive material found uncontrolled and outside the Restricted Area will be brought to the immediate attention of Radiation Protection supervision. Action will be taken to ensure that the radioactive material is surveyed, labeled and properly dispositioned (e.g., returned to the Restricted Area). The incident will be investigated by the Project Radiation Protection Manager and corrective action (s) initiated. 3.2.6.5 Storaae of Radioactive Material Implementing procedures will be developed to specifically address storage of radioactive materials. Interim radioactive material storage and liquid processing areas will require a safety evaluation to ensure compliance with NRC Generic Letter 81-38. Access to radioactive material storage areas located inside the Restricted Area will be controlled by Radiation Protection personnel. Storage and processing of materials will be consistent 1 with analyzed activities and radiation levels. Storage areas will ' be regularly surveyed and inventoried. Temporary radioactive material storage areas within- the Restricted Area will be designated by Radiation Protection personnel and posted in accordance with Section 3.2.5.4. Consideration for establishing temporary radioactive material storage areas will include:
- 1. The number of temporary storage areas and length of storage time in these areas will be minimized.
- 2. Radioactive material storage . areas and surrounding areas l will be surveyed on a routine basis.
l 3. Radioactive material will not be stored outside except for short periods during transit, or if packaged in accordance with D0T requirements while awaiting shipment. 3.2.6.6 Accountability and Inventory of Radioactive Material p l Implementing procedures will specifically address the accountability l of radioactive materials. Radioactive material storage areas will be l ' controlled and periodically surveyed. The status of radioactive material storage areas will be periodically reviewed and include: 3.2-51 l
i PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3
- 1. A list of storage areas to evaluate the continued need for the storage areas and/or the. materials and the types and radiological parameters of the areas (e.g., high radiation material, contaminated material, activated material, etc.).
- 2. Inspection of materials stored, to evaluate the status of materials and area c atrols (e.g., physical condition of containers, access control, posting, etc.).
Radioactive sources used for calibration, standardization and instrument checks will be recr.ived as indicated in Section 3.2.6.1. Implementing procedures will specifically address inventory and accountability requirements. Starces will be inventoried at least semiannually, taking into consideration: radioactive decay; comparison to Certificates of Calibration; and use and disposal of liquid and gaseous sources. The results will be documented and reported to Radiation Protection supervision. Scaled sources will be controlled in accordance with 10 CFR 30 through 34. Sealed sources will be leak tested with the frequencies and criteria for results based on activity level and type of source. The test will be documented and will be capable of detecting a minimum of 0.00S microcurie of contamination. If leakage is detected based on established criteria, the Project Radiation Protection Manager will be notified and reports will be submitted as indicated in Section 3.2.10. Sources will be maintained, locked or otherwise controlled when not in use. A source usage log will be maintained. If a source can not be accounted for during a periodic inventory or at any other time, actions will be taken to include the following:
- 1. Evaluate the physical and radiological characteristics of the missing source and evaluate potential hazards to radiation workers and the public.
- 2. Report the missing source as required by applicable implementing procedures and State and Federal regulations.
- 3. Initiate investigations to locate the source and/or to determine the reason for loss.
- 4. Prepare Radiological Occurrence report (s).
3.2.6.7 Special Controls Special circumstances may arise during the decommissioning of Fort St. Vrain that will require special handling considerations for radioactive materials. Examples of radioactive materials requiring , 3.2-52
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 special handling are radiography sources and highly radioactive waste (e.g., greater than 10 R/hr gamma radiation levels). Receipt of "special" types of materials / sources require the following actions:
- 1. Notification of the Project Radiation Protection Manager prior to bringing the radioactive material / source inside the Restricted Area.
- 2. Cavelopment and approval by the Project Radiation Protection Manager of a plan of action for control, safe storage and release of the radioactive material / source (e.g., use existing procedures, develop temporary procedures, etc.) prior to bringing it inside the Restricted Area.
- 3. Notification of the PSC Shift Supervisor prior to use of a source (e.g., radiography).
Special nuclear material, fissile material, and highly radioactive waste will be addressed on a case-by-case basis in accordance with administrative procedures and applicable regulations. Specific control, accountability and storage, if applicable, will be specified by the Project Radiation Protection Manager with the cognizance of the PSC Facility Support Manager. These details will be addressed by the Radiation Work Permit Program and/or temporary procedures and action plans, and will include the following considerations:
- 1. Use of special containers and/or shielding.
- 2. Use of special rigging and lifting devices for moving the radioactive material / source including any containers as applicable.
- 3. Temporary evacuation of non-essential personnel and suspension of activities in the area.
- 4. Temporary implementation of additicaal security measures.
- 5. Other special considerations for ALARA purposes as indicated in Section 3.5.1.
3.2.6.8 Release of Materials for Unrestricted Use Procedures for the release of materials and equipment from radiologically controlled areas will be developed. Material will not be released for unrestricted use if it contains detectable amounts of radioactive material. Instrumentation, counting times and survey techniques will be selected such that detection sensitivities are consistent with the applicable guidance of: i 3.2-53
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3
- 1. Contamination survev .or packages received and packages shipped to ensui. Snat they meet the D0T requirements for smearable contamination levels (e.g., external beta-gamma and alpha, and internal beta-gamma and alpha).
- 2. Radiation surveys (e.g., package contact, vehicle contact, specified distances from the package and the vehicle, and normally occupied positions in the vehicle cab) for the material and package and for the transport vehicle depending on the type of shipment (e.g., LSA, Exclusive Use LSA, etc.).
Additional elements described in the Radioactive Waste Manual implementing procedures will include, but not be limited to: Sorting and segregation of materials and processing to an acceptable form. Classification of the material in accordance with 10 CFR 61. Receipt survey of vehicles used to transport radioactive waste. Packaging, labeling and marking of material in accordance with 10 CFR, 49 CFR and Disposal Site Criteria. Shipment of material in accordance with 49 CFR and 10 CFR Disposal and off-site volume reduction arrangements. Additional details on radioactive waste management can be found in Section 3.3 of the Proposed Decommissioning Plan. 3.2.6.11 Badioactive Liouid and Gaseous Release Control liquid and gaseous effluent releases will be monitored and controlled using installed plant equipment and guidance provided in the Decommissioning Technical Specifications and Offsite Dose Calculation Manual. Typical- process controls will include: Sample and analyze the waste stream Calibrate and test instrumentation Monitor the waste stream Prepare a release permit Complete the release Record types and quantities of materi11(s) released i 3.2-55
l PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 3.2.7 Surveillance Radiological surveillance will be conducted routinely to identify radit. tion sources, determine radiological conditions, and comply with the requirements of 10 LFR 20. Special surveys will be scheduled by Radiation Protection Supervision as needed to evaluate radiological conditions in support of decommissioning activities. Elements of radiological surveillance program will consist of the following:
- 1. Routine surveys-general
- 2. Dose rate surveys
- 3. Surface contamination surveys
- 4. Airborne radioactivity surveys
- 5. Environmental sampling and analysis
- 6. Personnel contamination monitoring
- 7. Survey documentation and review Detailed radiological surveillance requirements and activities will be described in the Radiation Protection Manual administrative and implementing procedures. These procedures will specify the types of instrumentation, survey methods, and actions required when abnormal radiological conditions are discovered.
Calibrated instrumentation will be available for the detection and measurement of alpha, beta and gamma radiation. Air sampling equipment will be available for general area and breathing zone air samples. Survey and monitoring information will be used by the Radiation Protection staff and other support groups for:
- 1. Procedure development, as applicable
- 2. Decommissioning work packages development
- 3. Decommissioning engineering design criteria
- 4. RWP preparation
- 5. Radiation, contamination and airborne radioactivity tr< nd analysis
- 6. Pre-job ALARA planning
- 7. Pre and post job briefings Surveillance frequencies will be specified in implementina procedures with consideration given to hazards which may be encountered, potential for changing radiological conditions and frequency of occupation. Examples of surveys and associated frequencies include the following:
3.2-56
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3
- 1. Active work areas where radichNical conditions may change as a result of work being performed will normally be surveyed for radiation and contamination at least once per shift or more frequently if radiological coMitions could change (e.g., upon opening a radioactive system).
- 2. Active exit points from contaminated areas will be surveyed for contamination at least daily and once per shift during frequent use.
- 3. Eating areas used by individuals who have worked in radiologically controlled areas will be surveyed for contamination at least weekly.
- 4. Storage areas for solid radioactive waste and irradiated / contaminated components and equipment will be surveyed wet.H y when material and/or personnel have entered the area.
- 5. All- personnel and equipment exiting radiologically controlled areas will be monitored for contamination.
- 6. RWP's will normally specify the frequency of Radiation Protection technician coverage and surveys required (e.g.,
continuous, intermittent). 3.2.7.1 Routine Surveys s pey_3], Routine radiation, contamination and airborne radioactivity surveys will be performed to evaluate radiological conditions and verify radioactive materials are being adequately controlled. Survey data will be.used for-job evaluations, trend analysis, ALARA pre-planning and informing personnel' of radiological conditions. Action levels and associated responses will be established for abnormally high or L unusual survey results. Survey results will be made available to l workers entering radiologically controlled areas. Locations where routine surveys and monitoring will- be conducted include, but are not limited to:
- 1. Active work areas where conditions may change as a result of the work being performed.
i 2. Entry and exit points of contaminated areas (e.g. step off pads). l_
- 3. Offices, trailers, shops and trash receptacles outside radiologically controlled areas.
- 4. Major pathways inside radiologically controlled areas
- 5. Radiation Area boundaries for verification of posting adequacy.
- 6. Storage _ areas for radioactive wastes.
- 7. Entrances to locked High Radiation Areas.
- 8. Radiation Protection and Radiochemistry laboratories.
3.2-57
PROPOSED DECOMMISSIONING PLAN 6/28/91
-SECTION 3
- 9. Eating and break areas used by personnel who have been working in radiologically controlled areas.
- 10. Selected unrestricted areas, if appropriate.
Radiation Prvtection supervision will routinely review surveys with regard to necessity and frequency consistent with good radiological protection practices and regulatory requirements. Routine surveys will not normally be conducted in High Radiation Areas except as directed by Radiation Protection supervision. These surveys will be coupled with, or prior to, planned work activities l in those areas in order to maintain personnel exposure ALARA. 3.2.7.2 Dose Rate Surveys Dose rate surveys will be performed to provide specific radiological information on beta and gamma radiation dose rates. These surveys will normally be performed with portable, hand held servey instruments. Dose rate information may also be obtained through the use of fixed radiation monitors. Dose rate surveys will be performed to:
- 1. Assess changing radiological conditions in radiologically controlled areas which are frequently occupied.
- 2. Identify localized hot spots.
- 3. Provide data for pre-job ALARA planning and RWP preparation.
- 4. Establish " Low Dose Waiting Areas".
- 5. Monitor the receipt of radioactive materials.
- 6. Support the packaging / shipping of radioactive waste.
l 7. Ensure proper release of materials and equipment for l unrestricted use,
- j. 8. Monitor unanticipated spills or spread of radioactive l materials.
- 9. Assess radiological conditions during decommissioning work (e.g., breaching of a radioactive system, PCRV dismantlement etc.).
- 10. Provide data for environmental monitoring.
- 11. Support emergency response activities.
- 12. Establish and verify radiation area boundaries and postings.
- 13. Monitor laundered / decontaminated personnel protective equipment (e.g., protective clothing, respirators) prior to reuse.
3.2-58
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 Dose rate radiation survey instruments will be calibrated to the radiation (beta, gamma) being detected to assure an accurate, consistent, reliable and predictable response to radiation levels. 3.2.7.3 Surface Contamination Survevji Contamination surveys will be performed to provide specific radiological data on the levels of beta-gamma and alpha contamination, and will be performed to:
- 1. Monitor personnel and equipment exiting radiologically controlled areas to prevent the inadvertent release of radioactive material to uncontrolled areas.
- 2. Establish boundaries of contaminated areas.
- 3. Support the radioactive source accountability program (e.g., source leak tests).
- 4. Monitor the receipt of radioactive materials.
- 5. Support the radwaste packaging / shipping program.
- 6. Ensure the proper release of materials and equipment for unrestricted use.
- 7. Provide data for pre-job ALARA planning and RWP preparation.
- 8. Determine radiological conditions during coverage of jobs with changing radiological conditions (e.g., welding, grinding radioactive system opening).
- 9. Provide da'ta for environmental monitoring.
- 10. Support decommissioning emergency response,
- 11. Assess conditions following the discovery of a spill or spread of radioactive materials.
- 12. Survey TLD's prior to proceding for the prevention of cross contamination.
- 13. Detect and control " Hot Particles".
- 14. Monitor decontaminated personal protective equipment (e.g., respirators) prior to reuse.
- 15. Monitor applicable areas, such as clean waste dumps and landfills, salvage areas, warehouses, tool storage areas and contractor buildings.
- 16. Assist in personnel decontamiriation by monitoring for adequate decontamination techniques.
3.2.7.4 Airborne Radioactivity Surveys Airborne radioactivity will be measured in areas where personnel may be exposed to airborne particulates and tritium. Representative air sampling will be performed to provide measurements during work which has the potential for the generation of airborne radioactivity. 1 3.2-59 l
~ PROPOSED DECOMMISS10NillG PLAN G/28/91 SECTION 3 Continuous air monitors, breathing zone air samples and grab air samplers will be used for obtaining air samples. Air samplers that are expected to collect extremely high levels of activity will be exhausted back to their source. Airborne radioactivity surveys will typically be performed:
- 1. During work operations known or suspected to cause airborne radioactivity (e.g., grinding, welding, burning, cutting, hydrol azing, vacuuming, sweeping or use of compacting equipment).
- 2. During work that involves the breach of a radioactive system.
- 3. Upon initial entry and periodically thereafter into any area eown or suspected to contain airborne radioactivity concentrations in excess of 25 percent of DAC.
- 4. Immediately following the discovery of a significant spill oc spread of radioactive materials or whenever airborne radioactivity levels are suspected to have changed.
- 5. Periodically in radiologically controlled areas where the potential for airborne radioactivity exists.
- 6. Any time respiratory protection devices or DAC-hr accounting are used to control internal radiation exposure.
- 7. When continuous air monitoring is performed, high volume grab samples or breathing zone air samples will be periodically taken to verify that continuous air monitoring of the work area is representative of the breathing zone (This surveillance will also be performed forTritium).
- 8. Periodically to verify the effectiveness of the Respiratory Protection S cgram.
Air sample counting equipment will be available to measure alpha, beta and gamma emitting radionuclides. Collection efficiencies for air sampling media will be determined. The minir .. detectable activity (MDA) for air sampling equipment will be at least one (1) DAC in one (1) hour for the radicisotopes most likely to be present. Air sampling equipment will be calibrated using guidance in NRC Regulatory Guide 8.25 " Calibration and Error Limits of Air Sampling instruments for Total Volume of Air Sampled", (Ref. 32) August 1980. Periodic tests will be conducted to verify the adequacy of engineering controls that are usea to minimize airborne radioactivity. Typical controls that will bc verified are: 3.2 60
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 i Room vent 11ation systems ' Portable ventilation systems Air locks Ventilation hoods Containmentdevices(e.g., tents,gloveboxes,etc.) 3.2.7.5 Environmntal Samolina and Antilysil Environmental sampling and analysis will be conducted during decommissioning. The current PSC Radiological Environmental Monitoring Program (REMP) will be continued, en part, specifically tailored to determine the effect on radiological conditions of the environment due to decommissioning activities. In addition, the Offsite Dose Calculation Manual (0DCM) will provide the methodologies to assure compliance with fort St. Vrain Decomr.iissioning Technical Specifications related to liquid and gaseous radioactive ef fluents. This program will demonstrate compliance with 10 CFR 20, 10 CFR 50 Appendix A (GDE & 64) and Appendix ! and 40 CFR 190. Specific sample types and locations will be addressed in the REMP and ODCM. Typical environmental monitoring techniques that will be utilized, include: { ' Area TL0s Soil sample analysis Water sample analysis Vegetation sample analysis 3.2.7.6 Personnel Contamination Monitorina Adequate personnel contamination monitoring instrumentation will be available to control the spread of contamination and hot particles l to uncontrolled areas. Radiation Protection Manual implementing l procedures will require monitoring upon exiting radiologically controlled areas and will establish acceptable methods for performing personnel frisking. Whole body contamination monitors (e.g., Eberline PCM 18) will be used at radiological controlled area exits, as appropriate. Sensitivities of these instruments will be set to detect contamination levels equal to or better than conventional hand frisking methods (e.g., 5000 dpm/100cm2). In areas where whole-body contamination monitors are r:ot available, a hand-held frisker will be used to monitor personnel contamination. Typical frisking techniques that will be required include: 3.2 61
'*r- , * -- -- -.e .. __ .__________mm___.m_mm_. _ _ _ _ _ _ _ . _ _ _ _ _
PROPOSED DECOMMISS10f(ING PLAN 6/28/91 SEC110N 3
- 1. Maintain frisking speed of less than 2 inches per second.
- 2. Maintain detector to body distance at less than 1/2 inch.
- 3. Pause (approximately 5 seconds) at the nose and mouth area to check for indications of inhalation / ingestion of radioactive material.
- 4. Pay particular attention to feet (shoes), elbows, knees or other areas with a high potential for contamination..
- 5. Ensure a total frisking time of greater than 2 minutes to cover at least 10% of the body.
- 6. Maintain background for frisking at less than 300 cpm.
When background levels are unacceptable (e.g., greater than 300 cpm) for personnel frisking, actions will be taken that include one or more of the following:
- 1. Move the whole body frisker and/or the hand held frisker (and the contamination control point) to an area that has an acceptable background level (e.g., around the corner, behind a column, etc.).
- 2. Shield the frisking area and equipment to reduce background.
- 3. Frisk for gross contamination levels in the high background area, but locate the equipment for final frisking at a remote area, and provide contamination control for the passage to the remote frisking location.
3.2.7.7 Survey D2cumentation and Review Radiation Protection supervision will review completed survey documentation to ensure appropriate, adequate and complete information is recorded. The supervisor reviewing the survey will ensure that the recorded results are legible, in accordance with Radiation Protection Manual impicmenting procedures and consistent with anticipated levels and will determine the reason for any variances. Information that will typically be included on survey maps or forms is: Date and time of survey Location of survey A sketch or description of the area or component surveyed Instrument type and serial numbers instrument calibration due date Name and signature of surveyor The results of evaluations will be documented on approved survey forms which will be made available to personnel entering l 3.2-62
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 radiologically controlled areas. Survey data will contain enough detail to provide personnel with adequate information concerning radiological conditions existing in the area surveyed. Survey maps will include, as applicable:
- 1. Contact and general arers dose rates
- 2. Contamination levels
- 3. Airborne radioactivity levels, if applicable
- 4. Identification et specific hazards (i.e., hot spots)
- 5. Location of rar.lological boundaries Personnel contamination detected on hair or skin will be promptly removed under the supervision of trained Radiation Protection personnel. Personnel skin and clothing contaminations will be documented and evaluated to help improve contamination controls.
Personnel contamination forms will include such items as:
- 1. Names of individuals involved
- 2. Survey results
- 3. Decontamination methods
- 4. Results of decontamination
- 5. Area where contaminated
- 6. Areas worked
- 7. Radiation Work Permit number
- 8. Corrective action to help prevent recurrence of contamination Survey records will be filed and maintained so that previous radiological conditions can be determined. All original survey records will be maintained and retained in accordance with 10 CFR 20 and fort St. Vrain Decommissioning Technical Specifications Licensing Information, Section 5.6 Record Retention.
3.2.8 Instrumentation A sufficient inventory and variety of operable and calibrated portable, semi-portable and fixed radiological instrumentation will be maintained to allow for effective measurement and control- of radiation exposure and radioactive material and to provide back-up capability for inoperable equipment. Equipment will be appropriate to enable the assessment of sources of gamma, beta, alpha and neutron radiation including the capability to measure the range of dose rates and radioactivity concentrations expected. . installed process and effluent monitors will be set up and operated using Offsite Dose Calculation Manual implementing procedures. 3.2-63
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 Calibration prer.edures for process and effluent monitors will be
,slemented and m0intained.
Accuracy requiremer.ts, renote read-out utilization, alarm set points and conditions, and types of surveying or monitoring to be performed I will be specifie:I in Radiation Protection Manual implementing procedures. Remote and special monitorfng equiument will be obtained and calibrated per approved procedures if 'equired during the decommissioning. Counting instrumentation is located in multiple laboratory facilities providing back-up capability. Methods to perform manual calculations as a backup to computerized systems will be contained in implementing procedures. The Radiation Protection Manual will contain administrative and implementing procedures for the following activities:
- 1. Instrument inventory and control
- 2. Instrument calibration ,
- 3. Instrument operating procedures 3.2.8.1 Instrument Inventory and Control Radiation Protection Manual implementing procedures will ensure that instruments are calibrated at the required frequency, functioning properly, issued to appropriate _ personnel and returned when necessary. Special use or dedicated instruments will be marked to ensure they are not used for other purposes. Adequate instruments will be available for radiation surveillance and associated radiation protection' measurements, taking into consideration:
- 1. Number of personnel and numbers of separate work areas requiring surveillance.
- 2. Frequency and types of surveys or measurements required to l support decommissioning activities.
- 3. Allowance for repair and calit. rations.
- 4. Efforts to minimize delays in personnel access and egress from radiologically controlled areas.
- 5. Dedicated instruments (if any) that will be required for emergency response.
A minimum instrument inventory level will be established to ensure that decommissioning activities will not be limited due to inadequate survey capability. Table 3.2-1, " Typical Fort St. Vrain 3.2-64
i 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 Decommissioning Radiation Monitoring Instruments" lists typical equipment which will be available. Instruments that are broken or require calibration will be tagged out of service by Radiation Protection personnel. The out-of-service instruments will be separated from operable instruments and placed in a designated location until they can be repaired and/or calibrated. A whole body counter will be maintained onsite and will be capable of identifying approximately 10% of the maximum permissible organ or body burden from those gamma emitting isotopes likely to be encountered (e.g., C0 60, Cs-137). 3.2.8.2 Instrument Calibrat19.n Procedures for calibration and response checks of radiation monitoring equipment and air sampling equipment will be prepared consistent with guidance provided in the following documents:
- 1. NRC Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)", (Ref. 33).
- 2. NRC Regulatory Guide 8.25 " Calibration and Error Limits of Air Sampling Instruments for Total Volume of Air Sampled",
(Ref. 32).
- 3. ANSI N13.1-1969, "American National Standard Guide to Sampling Airborne Radioactive Material in Nuclear Facilities",(Ref.34).
- 4. ANSI N42.14-1978, " Calibration and Usage of Germanium Detectors for Measurement of Gamma-Ray Emission of Radionuclides", (Ref. 35).
- 5. ANSI N42.3-1969, "American National Standard and IEEE Standard Test Procedure for Geiger Mueller Counters",
(Ref. 36).
- 6. ANSI N320-1979, " Performance Specifications for Reactor Emergency Radiological Monitoring Instrumentation", (Ref.
37).
- 7. ANSI N323-1978, " Radiation Protection Instrumentation Test and Calibration", (Ref. 38).
- 8. ANSI /IEEE Std 325-1986, "lEEE Standard Test Procedures for Germanium Gamma-Ray Detectors", (Ref. 39).
The primary calibration frequency for commonly used portable radiation monitoring instruments and portable air sampling equipment will be every 6 months, after repairs or modifications or when malfunctions are suspected. Semi-portable and fixed instrumentation 4 3.? 65
l PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 will be calibrated at least annually, after repairs or when malfunctions are suspected. Instrument performance checks (source checks) will be conducted in accordance with ANSI N 323 as prescribed in Radiation Protection Manual implementing procedures. At least annually, a review of historical maintenance and calibration trends will be performed for each instrument type. The review will evaluate instrument performance, and the adequacy of calibration frequencies. Calibration procedures will typically include the following:
- 1. Instrument specification and limitations
- 2. frequency of calibration
- 3. Description of operating settings / parameters
- 4. Environmental limitation (if appropriate)
- 5. References (e.g., instruction manual s , other related procedures, regulatory guidance, etc.)
- 6. Required equipment for calibration (e.g., sources, tools, jigs, test equipment, etc.)
- 7. Applicable drawings and schematics
- 8. Calibration data forms (including as found/as-left settings, instrument and source identification, charts, etc.)
Laboratory analysis equipment will be calibrated using National Institute of Standards and Technology traceable sources of appropriate geometries and energies. The radiochemistry laboratory will participate in an interlaboratory cross check program. Documentation of calibrations and cross-checks will be maintained as quality records in accordance with the fort St. Vrain Quality Assurance Program. Audits of radiation monitoring and air sampling equipment will also be performed in accordance with the fort St. Vrain Quality Assurance Program. 3.2.8.3 Instrument _Operatina Procedures An operating procedure will be prepared for each type of instrument in use, including emergency and special use instruments. Different models of equipment from the same manufacturer with similar features and performance characteristics may be combined into a single procedure, if the operating characteristics arc essentially the same, (e.g., Eberline R0-2 and R0 2A). Functional checks of portable radiation monitoring equipment will normally be performed daily or prior to use on the scale (s) expected to be used. Each scale not function checked will be clearly labeled to prevent its use. 3.2-66
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 Operating procedures will typically include:
- 1. User responsibilities l
- 2. Instrument and detector description
- 3. User instructions (including battery check, meter zero, range identification, etc.) '
- 4. Precautions and limitations S. Identificatior of proper check sources and associated jigs
- 6. Performance o source check and/or operational checks 3.2.9 Review and (Lt To ensure the Radiation Protection Program is effectively implemented and maintained, an organized system of review and audits will be impicmented in accordance with the Quality Assurance Plan as defined in Section 7.0, " Decommissioning Quality Assurance Plan".
Reviews and audits will be conducted by various project organizations and include the following components:
- 1. Radiation Protection Self Assessments and Reviews
- 2. Radiation Protection Corporate Oversight and Reviews
- 3. Quality Assurance Group Audits 3.2.9.1 Radiation Protection Self_ Assessment and Review The Project Radiation Protection Manager will be responsible for the quality of work performed by Radiation Protection personnel. The Project Radiation Protection Operations Supervisor will review for adequacy and approve completed radiation survey documentation on a day-to-day basis.
In order to further assure the quality of the Radiation Protection program, Radiation Protection Supervisory Reviews will be planned and conducted by all Project Radiation Protection Supervisors (including the PSC and Project Radiation Protection Manager) on a routine basis. These self assessment / reviews will include in-plant walk downs to directly observe the effectiveness of the Radiation Protection Program including, but not limited to, the following: ( l. Radiation protection staff effectiveness i
- 2. f acilities and equipment allocation and use
- 3. Worker radiological work practices
- 4. Compliance with Radiation Protection procedures, policies and specifications
- 5. Compliance with Radiation Work Permit and ALARA programs 3.2-67
PROPOSED OECOMMISS10NING PLAll 6/28/91 SECT 10ft 3
- 6. Conformance with project goals such as person-Rem dose, radioactive waste minimization, etc.
Deficiencies and other findings will be documented and addressed in accordance with Section 3.2.10, Radiation Protection Performance Analysis. 3.2.9.2 Radiation Protection Cornprate Oversight and Revimf lhe PSC Radiation Protection Manager and the PSC llealth Physicists are responsible for overseeing the Radiation Protection Program to ensure proper implementation. Periodic reviews, audits and monitoring of the Radiation Protection Program will be performed to ensure the fo' lowing:
- 1. The Radiation Protection Manual and implementing procedures are adequate to meet the Radiation Protection Program as described in Decommissioning Plan.
- 2. The Radiation Protection Manual and implementing procedures are being followed.
- 3. The Radiation Protection Manual and impleinenting procedures are adequate to meet the applicable regulations.
- 4. The Radiation Protection Program objectives are met.
- 5. The Radiation Protection Program is being effectively implemented and maint:ined.
- 6. The work completed is in accordance with the fort St.
Vrain Project Quality Plan. 3.2.9.3 Quality Asjutance Group Audits The PSC Quality Assurance group and the Westinghouse Quality Assurance organizations will conduct planned audits, reviews and assessments of the Radiation Protection Program in accordance with the fort St. Vrain Project Quality Plan and in compliance with applicable items of 10 CFR 50 Appendix B. 3.2.10 Radiation Protection PerformaqqL Analysis The Rtdiation Protection staff will establish methods to identify radiological incidents and radiological deficiencies in order to determine root causes and correct errors that cause radiological performance problems. Detailed performance monitoring requirements and activities will be described in and inplemented by the Radiation 3.2-68
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 Protection Manual which will consist of administrative and implementing procedures. 3.2.10.1 Radiolgq1 cal Occurrence Reports Radiological Occurrence Reports will be classified as either a
" Deficiency" or an " Incident". Principle elements of the reporting program will include, but not be limited to:
- 1. A Radiological Occurrence Report may be completed by anyone identifying a radiological occurrence. The report will include pertinent information relating to the occurrence (e.g., date, time, individual reporting occurrence, location, observations,etc.).
- 2. Radiological Occurrence Reports will ' submitted to Radiation Protection Supervision fa . review, and classification.
- 3. The Project Radiation Protection Manager and the PSC Radiation Protection Manager will approve corrective actions and implement disciplinary actions, when applicabic for designated classes of occurrences.
- 4. A root cause evaluation for designated types and levels of occurrences will be used,as applicable, to determine the circumstances and causes of the event and to develop short-and long term corrective actions to prevent recurrence. The evaluations will be conducted by the cognizant first line. supervisors and managers with assistance from the Radiation Protection staff.
- 5. A tracking system for Radiological Occurrence Reports and corrective actions will be implemented. The reports will be trended and evaluated periodically to integrate lessons learned, licensee experience and experience from others into Radiation Protection program improvement. Records will be maintained in accordance with regulatory requirements and company procedures.
3.2.10.2 Radioloaical Deficiencies
-Radiological " Deficiencies" are occurrences involving poor radiological work practices with relatively minor consequences, but ,
require supervisory action for proper resolution. Examples of ' occurrences attributable to Radiological Deficiencies include, but are not limited to, the following:
- 1. Failure to comply with radiological posting.
- 2. failure to comply with Radiation Protection procedures.
3.2-69
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3
- 3. Lost dosimetry.
- 4. Improper frisking.
- 5. Personnel contamination instances above a designated level.
- 6. Poor radiological work practices.
- 7. Improper use of, or problems with, respiratory protection equipment.
- 8. Eating, drinking, smoking, or chewing in tha radiologically controlled area.
- 9. Failure to comply with Radiation Work Permit requirements.
- 10. Unnecessary generation of radioactive or mi od waste. .
- 11. Operation and maintenance of equipment in a radiologically 1 unsafe manner.
These types of occurrences will be evaluated for possible deficiencies in areas such as training, procedures, equipment and human performance. Appropriate corrective action and follow up will be required by the Project Radiation Protection Manager or designee. 3.2.10.3 Radioloaical Incidents Radiological " Incidents" are occurrences that have, or could have the potential for, violating federal Regulations and fort Saint Vrain Decommissioning Technical Specifications or involve a serious breakdown in the effectiveness of the Radiation Protection Program. Examples of occurrences that would be considered Radiological incidents include, but are not limited to the following:
- 1. . Radiation exposures exceeding federal limits.
- 2. . Radiation exposures exceeding administrative limits without previous authorization.
- 3. Radioactive body burdens in excess of 25t' Maximum Permissible Organ Burden.
- 4. Unplanned exposures. of individuals to airborne radioactivity in excess of 2 DAC hours per day or 10 DAC-hours in any seven consecutive days.
- 5. Significant spills or spread of radioactive materials that affect decommissioning activities.
- 6. - Lost radioactive materials or radioactive materials found in uncontrolled areas.
- 7. Flagrant violations of dosimetry procedures, such as failure to wear or improperly wearing required dosimetry,
- 8. Lack of or improper access control for High Radiation Areas.
- 9. Improperly posted areas, especially high radiation areas.
- 10. failure to_ follow instructions and "stop work" orders.
3.2-70
PROPOSED DECOMMISSIONING PLAtt 6/28/91 SECT 10!i 3
- 1. Presence of radioactive liquids.
- 2. Potential for high activity sources or contamination (PCRV disassembly).
- 3. Potential for creating Hot Particles (PCRV pool work).
- 4. Transient high radiation levels (e.g. PCRV disassembly).
- 5. Airborne radioactivity (e.g. opening a system).
- 6. Specific radiological evaluations upon occurrence of unusual events (e.g., notify Radiation Protection for radiation monitor alarms).
- 7. Reminder of the need to follow the Radiation Work Permit which may have special controls associated and therefore longer lead time for Radiation Protection preparation.
- 8. Precautions for work when systems are in unusual status or configuration.
- 9. Changing of system status which may affect iaJiological conditions (e.g. , starting and stopping the PCRV water purification system).
Engineering Controls Radiation Protection engineering controls considered during ALARA reviews and RWP preparation. Examples of engineering controls will include, but will not be limited to, the following: Temporary shielding Specialty and remote handling tools Contamination control containments HEPA ventilation systems Decontamination equipment and techniques
- Remote surveillance systems 3.2.11.2_fadiation Work Performance Methods Planning for radiological work is an essential element in assuring an efficient use of resources, maintaining control of radioactive material, and l'eeping worker exposure ALARA. An objective of planning will be to provide adequate time for each affected department to prepare for radiological work. Planning will typically include the following considerations:
Provide adequate time for work area preparation, including removal of hazards, in addition to radiological hazards, to provide a safe working environment.
- 2. Decontaminate work areas to increase worker efficiency.
- 3. Install systems to contain radioactive materials.
3.2-72
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3
- 4. Provide workers with specialized training and other identified ALARA requirements.
- 5. Incorporate previous experience on similar jobs.
3.2.11.3 L!se of Temograry Shieldina Temporary shielding will be used to reduce dose rate levels near
" hot spots" and in the general area where work is to be performed.
Determination as to the type and amount will be evaluated by Radiation Protection personnel. An impicmenting procedure will be used for the control and use of temporary shielding. Additional details on temporary shielding is covered in Section 3.2.5 Dose Controls. 3.2.11.4 Contamination Control Equipp_gni Contamination will be controlled by employing a variety of engineering controls including HEPA ventilation, enclosurcs, strippable paint, and area / component decontamination. Exampics of contamination control methods that will be used, include:
- 1. The PCRV will be filled with water to control radioactive particulates that would normally be released when handled in air.
- 2. Containment or enclosures of appropriate size, equipped with HEPA ventil t. tion, will be used as necessary to prevent the spread of contamination while contaminated graphite blocks and other components are being removed from the PCRV or otherwise handled.
- 3. A work platform will be installed on the PCRV after the PCRV head has been removed. The platform will be equipped with a HEPA-filtered ventilation system that will exhaust air from beneath the work platform. This airflow will minimize the spread of contamination.
- 4. A debris collection system will be used in concrete cutting operations to minimize the spread of contamination.
- 5. Strippable paint or other suitable enclosures will be applied to some radiologically clean components or areas to prevent cross-contamination.
Additional contamination control methods will be considered during job planning and work package review. Isolation containments may be used to minimize the spread of contamination if the surrounding work area is uncontaminated or is much cleaner than the work area. 3.2-73
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 Radtation Protection Manual implementing procedures will provide guidance on the application and use of contamination control equipment. Examples of equipment include, but are not limited to: llEPA ventilation llEPA vacuums Containments Strippable paint Glove bags Sheeting (e.g. plastic, horculite) 3.2.11.5 Wnrk. AreiLhntllation Portable ll[PA venttiation units will be used in work areas for the control of airborne contaminants during work activities that have the potential for airborne centamination generation (e.g., burning, welding, grinding). Periodic tests will be conducted to assure pressure gradients and air flows are from areas of low potential airborne contamination to areas of higher potential contamination. Containments or tents may also be used in conjunction with 11 EPA ventilation to control airborne contamination. Radiaticn protection impicmenting procedures will provide instruction on the use and control of portable llEPA ventilation units. 3.2.11.6 Dncontamination fr9 cesses Various decontamination processes will be employed during decommissioning to prevent the spread of contamination, minimize the potential for internal uptake and reduce the associated radiation icvels for ALARA purposes. Examples of decontamination processes which will be used, include: Scabbling Abrasive blasting llydrolazing Ultrasonic cleaning Strippable coatings llEPA vacuuming Chemical cleaning llands on decontamination 3.2.11.7 Liquid and Solid Xaste Procrisfag Process controls will be applied to the identification, collection, processing, packaging and disposal of radioactive waste to ensure compliance with state and federal applicable regulations. 3.2-74
_ _ _ _ _ __~ _ _ _ _ . __ . . _ __ _ _. _ ___ 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 1 Radioactive waste processing and controls are addressed in Section l 3.3. l 3.2.11.8 Control System for Contaminated Tools Storage areas and hot tool cribs will be identified for the project, and will be used for the storage of reusable contaminated tools, components, equipment and materials. This designated storage will help to prevent the spread of contamination and maintain radiation doses ALARA. Implementing procedures will include the control and i use of contaminated tools and equipment, l 1 3.2.11.9 Area Postina Supplemental postings will be used to provide additional information to workers. These postings will be used in conjunction with the required posting and may contain the following types of information/ instructions.
- 1. Contact Radiological Protection for Entry
- 2. Hot Particle Controls Area ,
- 3. frisk Hands and feet Prior to Exiting
- 4. Potentially Contaminated Area (e.g., in overheads)
- 5. Internal Contamination (e.g., inside electrical panels)
- 6. Neutron Dosimetry Required for Entry (e.g., Neutron Source Storage Area)
- 7. " Keep Out" Radiography in Progress
- 8. Radiation Work Permit Required for Entry 3.2.11.10 Descrintion and functions of Protective Clothina Protective clothing will be provided for personnel working in contaminated areas and will be required as soecified on RWPs.
Selection and use of protective clothing will be based on known and expected contamination levels in the work area, as well as the expected working conditions. Protective clothing requirements will be specified on Radiation Work Permits. To ensure the proper control and use of radiological protective clothing, they will be used as indicated by Radiation Work Permits. Instructions for the proper donning and removal of protective clothing will be addressed as part of Radiation Worker Trsining. Containers for the disposal of used protective clothing will normally be placed at the exits of contaminated areas. Used protective clothing will be treated as potentially contaminated and handled as such. 3.2-75
PROPOSED DECOMMISSIONING PLAN 6/28/91 SEC110N 3 Implementing procedures will be developed for the control and laundering of contaminated protective clothing. Reusable protective clothing will be used whenever possible. Laundry will be monitored for acceptable residual contamination levels prior to reuse. I 1 l 1
-)
i i'
-1 3.2-76 :
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 TABLE 3.2 1 TYPICAL FORT ST. VRAIN DECOMMISSIONING MONITORING INSTRUMENTS DO5tMETERS Manufacturet Model Name Range Sn.udeit; Radiatsun Detated DCA N/A Self reading dosameter 0-2(O mR 5 mR Gamrra DCA N/A Self-readirig donianter 0-1(Ko mR 25 mR Ga uis l'ORTABLE SL'RVEY ME1FRS Manufacturer Model Name Range hitivity Ra&auon Detected Eberlane RO2 Air sua chamber with 0-5 to 0 5000 mR'hr 0.5 mR/hr nunimum. Osmns ame rate, RO 2 A heta window, battery 0 5 to 0 50.000 mR!ht 60 kev to 3MeV beta eine rate powered + + 20 % Eberkne RO 5A/D Air ion chamber with 0.1 to 1999 mRths and 0.5 mR/hr mirumum. Osmina donc raie, beta usadow, battery 0.01 to 199.9 RN 80 kev to 3MeV beta Ame rate powered + 20 % Eberkne RO 7, Air son chamber with 0 to 20,000 R!bt Varies based on probe Gamtre ame rate, RO-7Bil, remou detector (to 500 beta &me rate RO 7LD, h1 with beta window, s RO-711f 1 hattary powered Eberkne E 500Bl Portable GM beta- 042 mR/hr to 0 05 mR4r trummum, Garrurs &me rate llP 177 gamma Ame rate sneter 0 2,000 mR/hr 40 kev to 1.3 MeV with internal and entert.al GM detectors Ltechne Teletator Portable high range Five norJanear ranges 80 kev to 3 Mev Gamma Ane raie 61128 gamma done rate to 1000 R/hr + 20 % utstrument wth entendtag probe and two energy cornpensated GM delators Eber!tne PNR 4 Neutrea Survey Meter Four ha log daade4 Bermal to 10 MeV Neutron Amo rate full scale readings of 5.50,500 and 5k mrem /hr l l l i
. - _ - . - m ,_ ..
PROPOSED DECOMMISSIONING PLAN 6/28/91 ; SECTION 3 TABLE 3.2-1 (Continued) TYPICAL FORT ST. VRAIN DECOMMISSIONING MONITORING INSTRUMENTS LOW LEVEL CONTAMINA110N ML1ERS Manufatturer Model home Range Se:Wtivity Radeauon Detated I Eberline E 5201 OM survey meter,1,4 0-24.(Ko t em Approarmately 64 for Deta f(P 190A to 2.0 mgicm' l' end l' Tc-99 platal window uncompensaimi mourt e GM deta: tor Ekrlum PAC 40 Alpha survey neter 0400.000 cpm 50% of 2pi for Alpha ordy 3/ Four duode Linh g datnbuted plated AC-21 (tm) noter with alpha alpha sourte gas flow purportumal detector Eherians PAC-45/ Alpha survey meter 02Mcpm 20 to 30% of 2ps for a Alpha only AC 3 Four dande Lia l.og dutnbuted plated (tm) meter with alpha - alpha source actotdledon deta-lar depending on window thnknesa LOW LEVEL DOSE RATE ML'TERS Manufacturer Model Name Hanse Seemadeity Radistmn Detweed Ludlum - Model 19 Portable gamma rate 0-23 to 0-5,'O nuero Energy dependent Osmme tmly Micro R noter with mternal Nel R/hr in 3 ranges Meter utntitletmn desator PERSONNEL fRISKT.RS Manufacturw Model Name Range Sendtjesty Radiation Detected Eherline RM14 fnnker, pos: cake OM, 0-300 to 040.000 cpm $000 dpm/100 cm' Beta (10%), gattute with alerta, AC or - (t 4) hattery operation t Eberline PCM 18 Contaimrution Morutor N/A 30lKk5000 Beta, pauna 8 _dptn/10lk m IRT PM Portal Morutor 160 700,(10 cptsi Vanos with Osmma background Derime RM 15 . Single thannel count 0 500 to 0 500,000 $000 dpm/100 cm8 Reta (10%), gamna rate awter, GM cpm (14) pancde probe, with alarm. AC or battery operstma P
,vm-,,. w- u .-,ren, < , . ...n ,,,eo,_n.. l- ,-. . , - - - , . . , , , - , - _ , , . + . . . , , ,. ,r v,e-.,,-v,, , , ,-,,,mg. r -
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 TABLE 3.2-1 (Continued) 1YPICAL FORT ST, VRAIN DECOMMI%IONING MONiiORING INSTRUMENTS AIR 54M't_l RS Manufacturer Model Name Range Semitivity Radiat4on l Detected 1 RADeco H 809VI Portab's AC pimered 3 cubic fue pre nunute N/A N/A air um,,ler MSA FIXT. Portable battery 2 hters per nunute N!A N/A FLO po*ered air sampler Eberline PINO I A Att Moruke 1010' cpm Ca 137 56pm ht for lieta pamculate 1 10" uCt /cm' 3 61 crmht for l a 10 " uC4/ t m' RA$P Low Air sampler 0.1 to 5 cha N/A N/A v olurne staple High Air umpler 10 cubic feet ivr N/A N'A Volume nunute AREA MONITORS Manufacturer Model Name Range Semilitity Radiation Detected Eberkne RM 16 or Area Monitor Vanable taned on 0.03 mR'hr Gamma done rate RM 20 pruhea 40 kev to 1,3 MeV with probes SAMPLE ANALYSIS AND CALIBRATION LABORATORY EQUIPMLNT __ Manufacturer Model Name Harshaw TASC 12 A6 Gas Flow Proportumal Tennelec LB5100 Gas flow Prtivrtional Eberkne BC-4 0 M Detector Cannstra Spectrum F 5enes Celt detector Bn kman 3801 Liquid 5ctnullsium Bakman Shadow Chair Sodiurn lodide Detectors or equi alent Yanous Vanous Air sampler cahbrators Ytetoreen 570 Condenser R Meters Vanous Vanous Radiatma nources with traceahnhty to NisT LL Shepherd Model 99-1 ShiclJed cahbratmo sources
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 INTENTIONALLY LEFT BLANK
l PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 i PSC RADIATION v[lilNSHOUS[ PR0f[Cil0N MAhAG[R PROJ[Cf DIRECTOR l l IECHNICAL PROJLCT PROJfCT CONTROL OP[ RATIONS
$[RVICES MANAG[R RADIAllDN MANAG[R MANAGER l 201[C110N MANAGER (1)
ALARA COMMIT 1(( l RADIAf10N RADWASTE ALARA 1[CHNICAL PR01[Cil0N SUPERVISOR (1) SUP[RVISOR(1) $UPPORT OPERATION SUPERVISOR (1) , $UPERVISOR(1) l l l l RADIAll0N PROIICil0N l IECHNICIANS (!!) Figure 3.2 1 Westinghouse Team Radiation Protection Organization Chart
I 6/28/91 PROPOSED DECOMMISS10filf4G PLAll SECT 10f4 3 RAtlA1104 P90t[Cfl0h PeoGRAu (PCP. 5ttilD4 3.2) RADIAtly4 agggggggg Pacittt]On p4Dittt10N mA4wAL yggggg,3
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6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 3.3 RADIDACTIVE WASTE MANAGEMENT This section addresses the technologies, equipment, and procedures ' to be implemented for the management of radioactive waste during the fort St. Vrain (FSV) decommissioning project. These technical , approaches are based upon experience and address facets of planning, ! decontamination, packaging, storage, transportation, volume reduction or beneficial reuse, and final disposition of the waste materials, while minimizing secondary wastes. In developing the Radioactive Waste Management Program, many , elements were considered, including the following:
- 1. End use of the facility
- 2. Location and availability of disposal facilities
- 3. Potential for offsite release during D/D operations
- 4. Preventing contamination of uncontaminated areas
- 5. llse of existing buildings to support the waste packaging operations
- 6. Methods of approach related to waste type, waste class, and impact on safety
- 7. Cost effectiveness
- 8. Logical approach to the D/D operations
- 9. Ensuring that the occupational exposures are maintained as low as reasonably achievable-(ALARA)
- 10. Minimizing the impact on the health and safety of the .
general public
- 11. Maintaining flexibility for waste management to allow for unexpected wastes and changes in available technology
- 12. Effective implementation of a Process Control Program for radioactive wastes This section contains a description of the following activities associated with the radioactive waste management program:
Spent fuel disposal (3.3.1) Radioactive Waste processing (3.3.2) _ Radioactive waste disposal (3.3.3) Disposal of non radioactive wastes (3.3.4) 3.3.1 Spent fuel Disposal Although not related- to proposed decommissioning plans, the
- following information is provided on the ultimate disposition of the fort St. Vrain spent fuel. The preferred plan to manage all Fort 3.3-1
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 St. Vrain spent fuel is to ship the spent fuel to a DOE facility in Idaho. The original ihree Party Agreement (Ref. 40) between PSC, General Atomic (GA) and the Atomic Energy Commission (AEC) provided storage for eight segments of fort St. Vrain spent fuel at the Idaho National Engineering Laboratory (INEL) facility. To date. PSC has shipped three segments of spent fuel to INEL as a result of three previous refuelings. In 1988, the agreement was modified (Ref. 41) to clarify the intent of the agreement and to assure the storage of five remaining spent fuel segments (approximately 1242 fuel elements) at DOE Idaho, pending reprocessing or transfer to the federal high level waste repository (HLWR). Under terms of the existing Three Party Agreement, PSC is responsible for the interim storage of one segment of fuel (approximately 240 fuel elements) until the HLWR is available, now estimated to be available to receive fort St. Vrain spent fuel in approximately 2020. The shipping schedule for the fort St. Vrain spent fuel was planned to commence early 1991 and be completed in late 1991. Based on DOE verbal commitments, the shipping schedule will include the acceptance of the ninth and final segment of fort St. Vrain spent fuel. Due to tht unc etain schedule for shipping of spent fuel to Idaho or other DOE facilities, PSC is pursuing an alternate plan to license, construct ind opt rate an Independent Spent fuel Storage Installation (ISFSI) in accordance with 10 CFR 72. To support this alternate plan, a 10 CFR 72 License Application was submitted in Reference 42. The ISFSI lacilicy is located immediately adjacent to the current site. The actual location is outside the plant's existing Protected Area, approximately 1500 feet northeast of the Reactor Building. The ISfSI, using thi. Modular Vault Dry Store (MVDS) System, is designed to store up to 1482 fuel elements, up to 37 MCRB's and up to 6 neutron sources. Utilizing the ISFSI alternative, defueling will be completed by mid 1992, which will allow decommissioning to commence 'ot lat r than mid 1992. 3.3.2 Radioactive Waste Processing 3.3.2.1 Proaram Description for materials that may contain licensed radioactive material, radiological surveys will be performed to determine the extent of the contamination or activation. Based on these results, options 3.3-2
1 I PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 for decontamination or disposal, packaging, and processing will be determined. Onsite packaging or processing of radioactive waste prior to transportation will be performed in areas appropriate for these activities. Examples of such areas are the Hot Service facility (HSF), the gas waste compressor rooms (Reactor Building level 1, el. 4740'), the Compactor Building, and the fuel Storage Building, items not considered for decontamination or items that, following decontamination, are considered to have too high a specific activity for offsite volume reduction, will be packaged and shipped directly for disposal at a licensed burial facility. Greater than Class C (GTCC) wastes, if any, will be packaged for onsite storage and subsequent shipment to a designated storage or disposal facility. Radioactive wastes are expected to be categorized as follows:
- 1. P_gisnLially cod 11minated or reouirina minor spal dg.qpalamination: These include potentially contaminated materials that: 1) appear to be uncontaminated; 2) all surfaces are easily accessible; and 3) have a small surface area-to-weight ratio will be surveyed to determine if the material can be released for unrestricted use without decontamination or with minor decontamination effort, for example, a small surface area with only spot and/or smearable contamination can easily be decontaminated by such means as wiping, grinding, or removing the hot spot.
- 2. Eentral contamination with accessible surfaces and a 1qw arca to-weighi ratio: Materials with readily accessible surfaces for purposes of surveying and decontamination, and that possess a low surface area-to-weight ratio may be shipped directly to a licensed offsite processing facility for decontamination of the surfaces and final disposition.
- 3. General cont aminat ion / inaccessible surfagfs/hich surface Arn -to wtight ratio: Smaller metallic scrap or metals with inaccessible surfaces for performing surveys (e.g.,
previously sheared material) will be assumed to be contaminated and be packaged for shipment for further - processing at a licensed facility or shipped directly to ! burial. 3.3-3
1 PROPOSED DECOMMISS10filliG PLA!1 6/28/91 SECT 10fi 3
- 4. Attivated: Activated materials and high specific activity materials (primarily concrete, metals and graphite components), will either be packaged and shipped direct for disposal or to a licensed f acility for further processing and volume reduction.
Radioactive materials as categorized above will be evaluated to determine the optimum method for release, decontamination, or shipment offsite for further processing or for burial. The following onsite and offsite methods will be considered.
- 1. Onsite processing of liquid wastes.
- 2. Onsite filtration of airborne wastes.
- 3. Onsite decontamination.
- 4. Onsite waste volume reduci. ion.
- 5. Onsite packaging.
- 6. Offsite decontamination.
- 7. Offsite volume reduction.
- 8. Offsite repackaging / consolidation for disposal.
3.3.2.2 OnliLv Processino of Liagid Wastes During the Fort St. Vrain decommissioning project, contaminated water will be generated through several processes (such as diamond wire cutting, flooding of the PCRV, rinsing of contaminated components removed from the PCRV) and through decontamination operations. Flooding the PCRV will put into solution radionuclidos that exist in the PCRV as a result of activation and plateout. Of = primary concern are tritium and the gamma emitting isotopes ts-137 ' and Co 60. Co-60, Cs 137 (particulates) and other radionuclides that could enter the PCRV shield water can be removed using the water system with ion exchange resins or particulate filters. The contaminated water will be purified to releasable levels as defined in 10 CFR 20 . by means of the disposable demineralization and filtration system, which is a part of the PCRV water cleanup and clarification system described in Section 2.3.3.6. Operating simplicity of this system will minimize the radwaste movement, handling and personnel exposure. Spent resins and filter media requiring stabilization will be orocessed in accordance with j the Process Control Program (PCP). The PCP shall contain the l current formulas, sampling, analyses, tests, and determinations to ' be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet ; 3.3-4
6/28/91 PROPOSED DECOMMISSIONING PLAN SEC110N 3 solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, 49 CFR 100, State regulations, disposal site burial requirements, and other requirements governing the disposal of solid radioactive waste. When possible, this will be done inside the disposal package or liner to minimize additional waste handling prior to disposal. In assessing levels of tritium in the graphite reflectors and side spacer blocks, the fort St. Vrain Activation Analysis (Appendix 11 af this plan) conservatively assumed upper limit concentrations of impurities in the graphite, and also assumed that none of the tritium formed by neutron activation of impurities migrated out of the graphite. Based on these assumptions, the amount of tritium residing in the graphite was computed to be approximately 100,000 Curies, llowever, it is expected that the actual amount of tritium will be significantly less than this. Based on data on measured tritium release rates from graphite and assuming the conservative estimate of 100,000 Curies, tritium levels in the water system were estimated as a function of time for various purge rates from the system. Since the tritium cannot be removed by mechanical means, the tritium will be removed by a feed and bleed method and released to ensure concentrations are below 10 CFR 20 limits. The outlet stream from the demineralizers will be routed to the cooling tower blowdown line, and makeup water will be added to the discharge stream, if required for dilution, prior to release to the environment. During unforeseen circumstances that preclude use of the normal discharge path, the water may be directed to interim short term storage until a controlled discharge can be performed. Tritium concentration will be determined by sampling and analysis. The maximum concentration in the PCRV water system is expected to be about 0.3 microCuries/cc, based on a maximum total water inventory of 310 Curies. During the initial purging operation, the PCRV water tritium concentration is expected to drop to below 0.1 microcuries/cc within 40 days, and continue to decrease thereafter. The integrated total tritium released from the graphite blocks into the PCRV water system is predicted to be slightly above 500 Curies. A plot of total tritium predicted to be in the PCRV water system vs. time is show in figure 3.3-1. Af ter one to two months of water cleanup and clarification system operation, the total tritium concentration in the PCRV is expected to ree h an equilibrium level well below the unrestricted release limit for water discharge (0.003 microCuries/cc). 3.3-5
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 3.3.2.3 Relenq_of Airborne Contamination Plant gaseous effluent filtration and monitoring systems will be operated and maintained as described in the Decommissioning Technical Specifications and the Offsite Dose Calculation Manual. The HEPA filter penetration and bypass acceptance @ in the Technical Specification surveillances are applicaf he 'J upon a HEPA filter efficiency of 96%. The HEF" fi1+ar ban, i be testea using the test procedure guidance in Regulatory Positwn C.5.a and C.S.c of Regulatory Guide 1.52 (Ref. 43), with a flow rate of at least 17,100 cfm to verify that the filter penetration and bypass leakage test acceptance criteria of 1% is met. The replacement frequency of the HEPA filters in the existing Reactor Building ventilation exhaust system is also identified in Decommissioning Technical Specifications, and is based upon either high exhaust radiation readings (or alarm) in the ventilation exhaust duct, or upon exceeding the maximum allowable pressure differential (which indicates that the filters are filled with dust). Effluent monitoring of the reactor building exhaust will be accomplished and reported using installed plant equipment and established procedures. Supplemental effluent air monitoring in the form of air samples for areas or operations remote from the Reactor Building with air discharge capabilities will be maintained. Monitoring capabilities include beta / gamma radiation measurement of sampl es. 3.3.2.4 Onsite Decontamination Technioues Onsite decontamination techniques will normally be used for processing and volume reduction of solid wastes. Solid wastes will be processed in accordance with written procedures. A general plan for solid waste processing is to initially identify the waste at the point of generatica as to the type of material and exposure rate, and segregate the material to allow for decontamination onsite or packaged for shipment to an offsite vendor for volume reduction or to an approved dispcsal site. Standard industry decontamination techniques will be used and may include the following:
- 1. Strionable Coatinas: Strippable coatings may be used to lift particulates from contaminated surfaces. A strippable 3.3-6
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 coating is applied " wet" to a surface in a manner similar to painting a surface. Additives in the coating are designed to attract and combine chemically with radioactive contaminants. Once the coating is dry, the contaminant is locked in the dried coating. The dried coating is easily " peeled" to allow stripping of the film containing the contamination. The stripped film can then be packaged and buried as a solid waste. Strippable coating may also be used to protect surfaces from becoming contaminated.
- 2. 01pmical or So_lvent Decontamination: Chemical decontamination is utilized principally for batches and is best used on a production basis for large volumes of similar materials, but may result in a hazardous radiologically contaminated mixed waste. Chemicals used for decontamination will be evaluated for hazardous constituents using 40 CFR and Material Safety Data Sheets (MSDS). Decontamination chemical wastes could possibly include acids, caustics, detergents and non-hazardous solvents. The specific chemical for a particular application will depend on the material to be decontaminated.
Acids or bases may be neutralized and solidified or used for water chemistry control in the PCRV water clean-up system. Detergents and other water based solvents will generally be associated with damp rags or wipes. If a mixed waste stream is identified, a treatability study will be performed to determine if it can be made non-hazardous. At this time, no mixed wastes have been identified. Furthermore, no processes are planned to be used during decommissioning thet will create a non-treatable mixed waste.
- 3. Drv Abrasive impinaement: Dry abrasive impingement (e.g.,
sandblasting) is effective for removing heavy or tightly adhering oxide films.
- 4. fixatives The application of fixatives may be used to fix transferable contamination prior to cutting or packaging.
- 5. Vacuum Cleaningt HEPA filtered vacuum cleaners may be used in areas of high gross transferable contamination.
3.3.2.5 Onlite Radioactive Waste Volume Minimization Project management and performance level personnel Will incorporate radioactive waste minimization practices into work procedures. Performance indicators will be developed to track total radioactive waste generated during decommissioning. The actual volume of waste generated for an evolution will be compared to the pre-job estimate 3.3-7
l PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 for that task. Radioactive waste volume reduction and minimization techniques discussed below will be used as appropriate.
- 1. Eersonnel Trainina Affected personnel will receive Radiation Worker training. This training will identify work techniques to prevent unnecessary contamination of areas and equipment, practices for reuse of materials, and policies to prevent the unnecessary generation of mixed or radioactive wastes.
- 2. Preventipn of Waste Unnecessary generation of radioactive and mixed wastes will be controlled by procedures established to evaluate and control chemicals brought onsite, and prevent unnecessary packaging, tool s and equipment from entering radiologically controlled areas.
- 3. Reuse of Materials Typical materials reused during the decommissioning include contaminated tools, equipment and clothing. Contaminated tool and equipment storage and issue areas will be maintained. Protective clothing and collection bags will be laundered, repaired and made available for reuse.
- 4. Searegathn and Packaaino Waste material after collection will be identified at the point of generation as to type of material, exposure rate and contamination levels, if known. At the segregation / packaging facilities, the waste will be further segregated as to form and expected end process Liquid wastes will be separated from solid wastes.
3.3.2 6 Onsite Waste Packaaina Radioactive waste packaging at Fort St. Vrain will be performed in areas that minimize radiation exposure to personnel, control the spread of contamination, and are adequate for packaging activities. Examples of potential onsite waste packaging areas are: Reactor Building refueling floor Hot Service Facility Compressor rooms (Reactor Bldg., El. 4740') Fuel Storage Building Temporary facilities designated for waste packaging Waste packages will be selected for each waste stream that meet the requirements for transportation and disposal. Examples of the waste containers that may be used are drums (52-gallon, 55-gallon), boxes (2'x4'x6', 4'x4'x6'), liners, high integrity containers (HIC's). , 1 3.3-8
a 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 sea / land containers, shielded casks, and other specialty containers.
- The capacity and weight limitations of each container are governed l by the activity levels, form and classification of the enclosed materials. The waste container to be used will be determined by the size, weight, classification, and activity level of the material to be packaged. Guidance for selection of appropriate packaging will be provided in radioactive waste procedures. In all cases, packaging selected will comply with requirements specified by 49 i CFR, 10 CFR 71, and the Disposal Facility Site Criteria, as applicable.
To the maximum extent practicable, voids in disposal containers will be filled with other decommissioning debris. This will reduce the total volume of waste for disposal. Therefore, since . voids in packages are filled with wastes that would otherwise be packaged separately for burial, a superior waste form is produced, efficiency is maximized, and project cost, disposal site allocation usage, and transportation risk are minimized. Alternatively, the onsite use of a mobile super compactor may be a cost effective means of volume L reduction. After appropriate waste segregation and packaging have occurred, the waste will be transported directly for disposal or transported to an offsite licensed facility for further processing i and final disposition. 3.3.2.7 Offsite Shipments of Radioactive Materials for Further l- Processina Cost benefit analyses will- be performed to determine if it is more cost' efficient _ to process . certain radioactive materials at an l offsite facility specializing in .the treatment of these materials, l Based on the results of_ these analyses, a significant amount of l - radioactive material generated during the decommissioning project l may be shipped to a licensed volume reduction facility. ! Methods described below are examples of volume reduction processes that may be employed. l- 1. Incinerable material may be transferred to a licensed l incinerator facility for burning. This may include such materials as paper, certain plastics, lubricating oils and solvents. When required by regulations, EPA characteristic tests (or other analyses) will be performed to verify acceptability of a material for incineration.
- 2. Low specific activity metals may be transferred to suitably licensed facilities for either melting and consolidation, or 3.3-9
-I l
l PROPOSED DECOMMISSIONING PLAN 6/28/91 ! SECTION-3 '1 l
- decontamination and release. A variety _of decontamination i options exist. including- abrasive (grit blasting), chemical and ultrasonic cleaning methods, a
-3. Volume reduction by compacting or super-compacting.
Waste packages;sent to offsite facilities will primarily be sea / land containers selected to meet the requirements of transportation and receipt at the coffsite processing facility. Voids in transport containers ~are not a critical concern. However, efficient management- of_ transportation resources will be an important consideration to minimize project- costs and reduce the total-- number of. shipments made ~ Only radioactive materials that are acceptable according -_ to the individual license (s) of the receiving facility > , will'be_ transported to that offsite processing facility.
. Radioactive material control and accountability procedures to
- accurately track material originating from Fort _St. Vrain during-receipt, sorting, processing, and - packaging for disposal .will. be developediand implamented. Only offsite processing facilities which provide.-adequate radioactive - material control - and - accountability procedures. will be selected to J perform decontamination, volume reduction or waste processing services.
3.3.3. . Radioactive Waste Disposal 3.3.3.I'Proaram' Description The radi_oactive waste _d_isposal program -'will- follow:10' CFR 20 and 10 CFR 61,? the-disposal-site criteria, and other applicable Federal and ' l _' State . regulations. - Radioactive waste processing, .' packaging, and shipping activities at- Fort -St. Vrain _will be performed..in
, accordance=with written procedures, Similar operations performed at .
-offsite facilities will be controlled as -di.rected by local L- requirements and specific facility licenses._- Radioactive waste may L
bei stored onsite, ' subject - to ' approved safety evaluations, storage 4 and separation Lcriteria- established in - Section -3.4 and 10, and
- , fapplicable State or NRC guidance.
GTCC.-waste,--if any, will be stored in the adjacent .ISFSI or -in a s+ructure which meets the design requirements to handle GTCC waste. . Ti.e waste will be stored until such time as it can be transported to- . a facility: licensed to accept-the GTCC waste, p !~ 3.3-10 w, - . - . . . , , -. - . . . . _ _ . .___ ____ _ _ _ __ _ _ _____ -
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 3.3.3.2 PIgjected Radiq1ctive Waste Generation Tables 3.3-1 and 3.3-2 identify the radioactive wastes that may be shipped for further processing. The pre-volume reduction totals and estimated number of waste containers are delineated on Tables 3.3-3 and 3.3-4. The initial estimate of the processed and volume reduced radioactively contaminated waste for disposal is 100,072 cubic feet, with 99,219 cubic feet from the PCRV and associated operations, and 853 cubic feet from the balance of plant (B0P). The waste from the PCRV consists of activated concrete, graphite blocks, other activated components, miscellaneous equipment and piping, and concrete rubble. PCRV waste is contaminated principally with Fe-55, tritium, and Co-60. The waste frnm the B0P consists of tanks, pumps, HVAC filters, and miscellaneous equipment and piping. There may also be radioactively contaminated asbestos. Af ter processing and volume reduction, it is estimated that the volume of radioactive waste will be segregated into the following categories: Class Volume (cubic feet) A 70,768 B 28,293 C 1,011 Due to uncertainties in the analysis, as much as 400 cubic feet of Class C wastes may be reclassified as GTCC. Waste volume estimates will change as ongoing planning and decommissioning operations proceed. 3.3.3.3 []3ssification of Radioactive Wastes Classification of radioactive waste is required by 10 CFR 20,10 CFR 61, and disposal site requirements. A Waste classification compliance program will be developed and implemented to assure proper classification of waste for disposal. This program will ensure that a realistic representation of the distribution of radionuclides in waste is known and that waste classification is performed in a consistent manner. Any of the following basic methods, used individually or in combination, will be used to achieve this goal: materials accountability (including process knowledge and activation analysis), classification by source, gross radioactivity measurements, and measuremant of specific radionuclides. 3.3-11
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 3.3.3.4 Transportation Plan Packages and pa .kaging for radioactive materials and waste will meet all applicable rigulations and requirements. Before packaging waste for shipment from Fort St. Vrain, each package will be i.'spected to to ensure it meets all applicable design and/or certifl cation requirements and that it is not damaged or impaired. A bar code capable of being read by computerized scanners will typically be affixed to the container and the corresponding lid as an aid to inventory and to track individual containers. The majority of radioactive material and waste shipments performed during the decommissioning project will be done by truck, in some cases, approved shielded casks will be employed due to radiation level s or limits for quantities of radioactivity in a package. Tables 3.3-3 and 3.3-4 provide preliminary estimates for the number of packcges anticipated during the decommissioning project. Tables 3.3-5 and 3.3-6 provide preliminary estimates for the types of packages anticipated during the decommissioning project. Special shielded shipping containers may be used for the steam generator primary a ssembl ies. The removal process and shipping container are described in Section 2.3 of this pl an. It is anticipated that the shipping container with the steam generator and grout will be shipped by rail for disposal at the Richland burial site. Transportation surveys and documents will be prepared prior to any shipment offsite. lo determine isotopic inventory and concentration for classification, onsite personnel will assess each loaded shipping container prior to transport. The actual routing of shipments may vary with weather and highway conditions. Additionally, local and state restrictions pertaining to radioactive material transport may affect some route selections, particularly in congested metropolitan areas. The carrier is responsible for selecting the appropriate route, which must conform to applicable federal, state, and local shipping, in accordance with D0T and NRC regulations. 3.3.3.5 Mixed Waste Contingency Plan Except for lead shielding, no sources of mixed waste are known to exist onsite. No chemicals or other substances are anticipated to 3.3-12
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 be used during decommissioning operations that may become hazardous wastes. It will be necessary for project management to authorize the use of any chemical or other substance that may become hazardous waste, if mixed waste is identified, it will be classified and stored onsite until regulations allow declassification or disposition. If mixed wastes are generated, they will be managed according to Subtitle C of RCRA to the extent it is not inconsistent with NRC handling, storage and transportation regulations. If technology, resources and approved processes are available, PSC and the Westinghouse Team will evaluate the processes for rendering mixed waste "non-hazardous a t.o determine its adaptability to Fort St. Vrain decommissioning activities. PSC does not intend to petition the EPA to delist any mixed waste. However, if PSC deteimines it is necessary to delist any mixed waste, the procedures outlined in 40 CFR 260,20 and 260.22 will be used to exclude that waste form from regulations. 3.3.3.6 Waste Storage Facilities Waste storage facilities planned for use during decommissioning activities include:
- 1. The Independent Spent Fuel Storage Installation (ISFSI) may be used for greater than Class C wastes (GTCC), if any, pending approval of an appropriate disposal site. (No GTCC wastes are currently expected.)
2 The New Fuel Storage Building may be used as a processing and storage area for dry low level wastes.
- 3. The Compactor Building may be used as a processing and storage area for dry and dewatered low level wastes.
- 4. The Reactor Building may be used for the storage of liquid and solid wastes.
- 5. Trailers and sea / land containers may be stored and used onsite to house dry and solidified low level waste.
- 6. Selected yard areas may be used for short term storage of packaged waste staged for transport.
The activity levals of wastes stored in these areas will be controlled to levd t as evaluated in an accident analysis. Safety evaluations have been performed that assess and permit storage of low level radioactive waste on the Fort St, Vrain site consistent with the guidelines of NRC Generic letter 81-38 (Ref. 44) 3.3-13 1
PROPOSEDDECONMISS10NINGPLAN 6/28/91 SECTION 3 and the Standard Review Plan (NUREG-0800)', Appendix ll.4-A (Ref. 45).- The Fort St. Vrain Technical Specifications permit possession and use of_ byproduct, source, and special nuclear material in quantities as required pursuant to 10 CFR 30, 40 and 70. Due to the building seismicity and other drainage and collection requirements for the storage of wet radwaste, PSC does not intend to ! store wet / liquid radwaste outside- the Reactor Building. The Reactor Building _ was designed and built with drainage- systems that route spillage to collection _ -points / sumps that are monitored for radioactivity and properly ;,rocessed. Other forms of radwaste may also be stored in the- Reactor Building without significant concern, due to the building's additional _ features relative to fire detection and suppression, and its filtered ventilation system, s The Compactor. Building is a steel building constructed on a concrete foundation, wl_th its own " wet pipe" fire suppression _ and fire- -1 detection systems. _ This building has two concrete basins that may ;
-~
be used to store barrels of dewatered wastes, consistent with the recommendation of NRC Generic Letter 81-38 (Ref 44). Other dry and solidified wastes may be stored in this building in amounts
' consistent with limitations of the decommissioning accident analyses. A radwaste compactor, with a self-contained llEPA-filtered ventilation system, is also housed in this building.
The- New- Fuel - StorageiBuilding willo also be:used to store packaged
. dry and solidified low-level radwaste. A safety evaluation has
-determined thatE no increase int an accident probability will result
~
from iradwaste storage in this location. As stated- in the decommissioning; accident analysis, fire suppression and fire detection systems will be -provided before _ combustible radwaste can
- be stored in the New Fuel Storage Building, e Trailers .and _ sea / land containers have been evaluated to house dry and. solidified radwaste. Accident scenarios have been postulated and 4
-the total- allowable activity levels for storage are controlled-accordingly. . Yard fire hydrants-are available for use-if necessary.
, Certain: large _ _ radioactive components (such as- helium circulators
- packaged for shipment) may be stored outside within _the protected area while - awaiting shipment offsite. - Tie-down_ systems will- be.
L considered 'for ' components stored- outside, and will be installed when needed. Steps ' will .be taken to protect - containers from external corrosion as required. 3.3-14 l- - - - . - - - - _. _.. . .. - . - - .. - -, , . .
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 3.3.4 Disposal of Non-Radioactive Waste Non-radioactive wastes will be disposed of by release to appropriate disposal facilities such as land fills,- scrap yards and scrap recovery facilities. Materials that are inappropriate for surface surveys, such as resin fines, will be sampled and appropriately analyzed. Materials found to be non-contaminated will be disposed of as non-radioactive waste, F L 3.3-15
l PROPOSED DECOMMISS10tilNG PLAN 6/28/91 i SECTION 3 INTENTIONALLY LEFT BLANK l 3.3-16
1 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 Table 3.3-3
-PCRV WASTE VOLUME ESTIMATES 11EM/$Y5 TEM CLAS$ CbC1 LSA NLMBER VOLLME (FT3)** C0hAINR$
Region constraint device & pin C 70 No 84 235 2 Metal controt rod reflectors C 3000 No 37 401 3 Metal block non control rod B 300 No 276 2025 13 Defueling blocks A Contam. Yes 1482 7200 75 Top Reflector graphite blocks A 0.5 No 1115 1515 8 Bottom reflector graphite blocks A 0.5 No 1215 1414 8 Radial reflector perm. & rmvbte A 0.5 No 480 1903 9 Large reflector blocks B <30 No 312 12600 50 1/2-size reflector blocks B <30 No 312 2160 8 Upper reflector keys (carbon steel) A 0.1 No i 24 192 2 Side Spacer blocks w/ boron rodt 30 1152 boron rods B 60 No 309792 Blocks w/ rods renoved A <0.1 No 1952 2393 10 bottom ref, blocks w/Nastelloy cans 300 2 76 Hastelloy C 10000 No 20061 Blocks with cans removed A 0.5 276 816 14 Lower reflector keys (Hastelloy) Core stoport blocks 8 A 1000
<0.1 No Yes 24 61 180 1468 1_
15 Core stw ort posts A <0.1 Yes 183 1 74 2 Core stoport floce colmns A <0.1 Yes 12 636 7 Misc. steel f rcm tecath CSF A <0.1 Yes 900 10 Metal on targe side reflector A <0.1 Yes 24 % 1 Core barrel A .02 Yes 1 1400 31 Lower plentro insulation A <0.001 Yes 940 10 Silica blocks (25,000 lbs.) A 0.5 Yes 503 12 Concrete top A <0.2 Yes 3744 9 Concrete - CSF A <.015 Yes 6240 15 Concrete side A <.005 Yes 18720 45 Concrete rthble Jackhanmer A <.015 Yes 706 16 Misc. Inconel parts on CSF A 0.4 - No 415 5 Concrete cutting debris - top A <0.2 Yes 210 Concrete cutting debris CSF A <.015 Yes 200 8 Concrete cuttino debris side A c.005 Yes 325 Helim purif ters in PCRV head A 0.15 Yes 10 480 5 Heti m diffusers A <0.1 Yes 4 1752 4 Helim circ shutof f velve assernbty A <0.1 Yes 4 192 2 Helim bellows A <0.1 Yes 12 1560 12 Steam generators A 2 Yes 12 20736 12 Thermocouples and Guide tubes B 50 No 105 1 Lower floor / appurtenances A 4 .01 Yes 1200 42 Platform / handling tools / Jib cranes A <0.01 Yes 576 6 Crane cable /drm/3 cucket inverters A <0.01 Yes 512 5 Misc. Containers A <0.01 Yes 288 3 PCRV Water System A <0.01 Yes 2030 2 Resins solidify, ship, bury A* 15 No 20 2720 20 Misc. soft waste A <0.01 Yes 12000 125 j PCRV TOTALS 113,972 (28 Estimated Burial Class - Specific burial class identification may require additional analysis in accordance with 10 CFR 61 Pre-volume reduced quantity.
PROPOSED DECOMMISSIONING PIAN 6/28/91 SECTION 3 Table 3.3 4 B0P WASTE VOL11ME ESTIMATES ITEM /SYSYEM CLAS CNkCT LSA NtMBE R VOLLME (F13)* C Al R$ Reactor isolation valves A <0.01 Yes 5 960 10 Refueling sleeves A <0.01 Yes 2 192 2 Sand from fuel storage wells A <0.01 Yes 750 Note 1 Sand f rom ecpipent storage wells A <0.01 Yes 225 Note 1 sand f rom Hellun regeneration pit A <0.01 Y es 135 Note 1 Auxiliary transfer cask sard A <0.01 Yes 15 Note 1 Hot cell facl(ity A <0.01 Yes 384 4 Sard f rmi hot cell f acility A <0.01 Yes 500 Note 1 Core sumort vent filters A <0.01 Yes 15 2 Caseous waste burge tanks A <0.01 Yes 2 2646 2 Liquid drain tank A <0.01 Yes 1 20 1 Gaseous waste vectun tank A <0.01 Yes 1 980 1 Gaseous waste empressors A <0.01 Yes 2 2058 2 Liquid waste monitor tank A <0.01 Yes 1 576 1 Liquid waste demineralizers A <0.01 Yes 2 192 2 Liquid waste receivers A <0.01 Yes 2 1152 2
- Iquid waste surp (sard) A <0.01 Yes 23 Note 1 Liquid trensfer pumps A <0.01 Yes 2 % 1 Liquid waste suno puros A <0.01 Yes 2 5 Note 2 Liquid waste filters A <0.01 Yes 2 15 2 Decon solution tank A <0.01 Yes 1 366 1 cecon recycle purp A <0.01 Yes 1 2 Note 3 Decun chem sumty putp a <0.01 Yes 1 2 Note 3 Purified Hellun filters A <0.01 Yes 2 14 Note 3 Hellun removat fitter A <0.01 Yes 1 96 1 Hellun getter uilts A <0.01 Yes 2 4 Note 4 HVAC filters A <0.01 Yes 1030 1 Fuel handling snachine A <0.01 Yes 192 2 Fuct handling machine sand A <0.01 Yes 420 Note 1 Small ard targe bore piping A <0.01 Yes 576 6 Reactor Bldg Drain System A <0.01 Yes 125 1 instrunentation & contrnLa A <0.01 Yes 225 2 TOTALS 13,991 46
* - Pre-volume reduced quantities.
Notes: 1- Vill be used as overfill 2- Vill be packaged with liquid transfer pumps 3- Will be packaged with Decon solution tank 4- Vill be packaged with helium removal filter l l
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 Curies 350 300 -- 250 - . . 200 -- - 150 - - 100 .- - - . . 50 -- O' O 10 20 30 40 50 60 70 80 Days After Flooding PCRV Curies Figure 3.3-1 Estimated Tritium Inventory in PCRV Water System
-PROPOSED DECOMMISSIONING PLAN 6/28/91 SECT 10N 3-INTENTIONALLY LEFT BLANK I:
i
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 3.4 ACCIDENT ANALYSIS 3.4.1 Introduction and Description of Decommissionina Acc;idents The purpose of this section is to evaluate the impact of potential Fort St. Vrain decommissioning accidents on the health and safety of the public. The ectivities, equipment and circumstances associated with decommissioning are different from those evaluated in the Fort St. Vrain Final Safety Analysis Report (Ref. 46) for power operations and refueling. Therefore, accidents analyzed for decommissioning are different from those evaluated for power operations and refueling. The risk of accidents resulting in a radiological release during decommissioning activities is considerably less than during plant operation, due to the removal of irradiated fuel from the Reactor Building. Since the reacter will be defueled prior to the commencement of decommissioning operations and all fuel will be removed from the Reactor Building, only non-reactor accident scenarios will be evaluated in this section. The focus of these decommissioning accident analyses will be on public health and safety. The following postulated accident scenarios have been analyzed, considering activation levels and isotopic compocition of components to be processed, and the anticipated dismantling activities:
- 1. Dropping of contaminated concrete rubble 2, Conversion construction near PCRV dismantlement
- 3. Heavy load drop
- 4. Fire S. Loss of PCRV shielding water
- 6. Loss of Power
- 7. Natural disasters The components with the highest activation levels were used in the accident analyses. Therefore. accidents that were analyzed bound the radiological consequences 'com other postulated accident scenarios. In evaluating the postulated accidents, conservative assumptions were made when data or knowledge to support more realistic analyses were lacking. Conservatism in this context is defir,ed to mean that the radiological consequences from the postulated accidents will be overestimated rather than underestimated.
i, 3.4-1
1 PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 A frequency-consequence diagram (Ref. 47) is shown in Figure 3.4-1; this figure defines three regulatory regions which are bounded by consequence limits established in 10 CFR 50 Appendix I,10 CFR 100, and the EPA Protective Action Guidelines (Ref. 48). Postulated accidents are assigned to one of the regions on the basis of their predicted consequences. The three regions are defined as follows:
- 1. Anticipated Operational Occurrences (A00s) - events that are expected to occur once or more in a plant's lifetime and whose dose consequences are analyzed in the plant's Safety Analysis Report (SAR) to demonstrate compliance with 10 CFR 50 Appendix I criteria.
- 2. Design Basis Events - events which are not expected to occur in the lifetime of the plant but may occur in - a large population of plants. These events are analyzed in the SAR to demonstrate compliance with criteria established in 10 CFR 100.
- 3. Emergency Planning Basis Events - events that are not expected to occur in the lifetime of most plants. The consequences of these events are analyzed in the EPA Protective Action Guidelines (Ref. 48) to establish criteria for emergency planning and environmental protection assessments.
As shown in Figure 3.4-1, the decommissioning accidents or events analyzed in this section are generally calculated to fall in the regions of the Design Basis and Emergency Planning Basis Events, due to the relatively low probability of the decommissioning accidents. A capsule 1ummary of the accident scenarios is given in Table 3.4-1. A sunnary of postulated accident dose consequences is presented in Figure 3.4-2.- The doses to an individual located at the decommissioning Emergency Planning Zone (a minimum of 100 meters from the Reactor Building, the Fuel Storage Building, and the Radioactive Waste Compactor Building, as defined in Section 9 of this plan) from these scenarios are presented in Table 3.4-2. From this table, the lioiting accident is a fire resulting in a whole body dose of 121 mkem and a dose of 215- mrem to the organ (lung). l These doses are well within the 25 Rem whole body dose and 300 Rem to any specific organ guidelines established by 10 CFR 100. These-doses are also a small fraction of the one Rem whole body dose and
'five Rem to any specific organ guidelines cited in EPA Protective Action Guidelines (Ref. 48).
3.4-2
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 The following natural disasters were considered in the accident analyses and are discussed in Section 3.4.9, i l External Event, Mitiaatino Feature Radioloaical Constquence Earthquake Low Probability Not postulated; See of Occurrence Section 3.4.9 High Winds, Bounded by lornado See analysis in Hail Section 3.4.9 ) Rainfall, Flood Site Location No release : Range Fire Plant buffer No release l The activity concentrations of the various components used in the following accident analyses were derived from the detailed neutron activation -analysis (Ref. 2), described in Section 3.1.4 and provided in Appendix II. Where chemical impurities were involved in neutron activation reactions, the maximum impurity levels permitted by the pertinent specifications were conservatively assumed to exist. With the exception of tritium concentrations, the radioisotope concentrations of interest used in the accident analyses have been taken directly from the activation analysis. Tritium concentrations predicted by the activation analysis were considered extremely unrealistic for the following reasons:
- 1. In the activation analysis, the dominant source of tritium was frota . activation of lithium impurities. The activation
- analysis assumed that no tritium fermed by lithium activation migrated out of the graphite into the primary coolant. The lithium concentration assumed to be present prior to irradiation in the graphite blocks was based on the maximum concentration permitted by the specifications. In actuality, lithium is relatively volatile and tends to migrate out .of the graphite during the high temperature graphitization process. Therefore, it is considered probable that the lithium impurity concentrations in the graphite used to form the large side reflectors and side spacer blocks were an order of magnitude lower than the maximum specification limit.
- 2. The large graphite side reflectors and side spacer blocks were exposed to relatively low temperatures (300-500 degrees c
C) during reactor operations. These low temperatures preclude a significant amount of tritium from being i l 3.4-3
]
l 1 \
~
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION-3 chemically absorbed in the graphite and retained. Since tritium has a small atomic radius, it is likely that tritium formed by activation of lithium (Li-6-and Li-7) will migrate out of the graphite. Due to this temperature dependence of chemical absorption, it is considered that tritium ' concentrations -are two or three orders of magnitude below those ~ predicted by the activation analysis.
.3. In the presence of moisture, hydrogen atoms from water molecules compete with and replace tritium atoms at active carbon sites-in the graphite matrix, releasing tritium from the' graphite. 'Before the graphite blocks-- are removed from the- PCRV, they will be submerged under water when the PCRV.
is flooded, which is expected to result in the release of a substantial fraction of tritium. l -
. Based-on the- effects- noted above, it 'is considered that a value of 10 uti/g of tritium represents a- conservative estimate of tritium concentration in the-large side reflector and side spacer blocks (Ref.'49). . While'this-concentration is.-a factor of approximately 40 u
below that projected -in'the activation analysis for these blocks, .it-provides. a more _ realistic representation of the- tritium concentration of the graphite blocks after they are removed from the PCRV. Therefore,. a tritium concentration of.10 uti/g in the -large side reflector and side spacer blocks is assumed for the postulated
- decommissioning accident scenarios.
3.4.2' Assumotions The following -are the major assumptions. used in the analysis' of postulated accidents which may.-occur during: the dismantling-activities: [. - L 1. The reactor is defueled and all irradiated ' fuel is removed from the Reactor Building. .
- 2. Since all feel is removed from the reactor, there will be no
-need . for shutdown /cooldown systems such as decay heat removal.
3 '. : The Reactor Building ventilation system will remain-toperable, providing flitration of ' effluents to the environment, while the potential exists for drop of a large f- activated' graphite block. 3.4-4 l l I
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3
- 4. The analyses for some of the ..ccidents conservatively assume a Curie content that exceeds allowable Curie contents for a low Specific Activity (LSA) Type A-2 waste container, as specified in Table A-1 of 10 CFR 71.
- 5. A worst case atmosphoric dispersion factor of 3.53 E-2 sec/m3 has been calculated and is used in the accident analyses, with the exception of the tornado accident, which utilizes an atmospheric dispersion factor of 4.59 E4 sec/m3 These atmospheric dispersion factors were calculated using the guidelines presented in Regulatory Guide 1.145 (Ref. 50) and are based on a minimum distance to the decommissioning EP2 of 100 meters. The atmospheric d,spersion factor of 4.59 E-4 sec/m3 represents the annual average dispersion factor for fort St. Vrain, and is considered to be conservative in the event of a tornado.
- 6. All releases to the environment are assumed to be ground level releases.
3.4.3 QC2pfina of Contaminated Concrete Rubble Accident 3.4.3.1 Identification of Quus After the majority of the PCRV top head concrete is removed in large pieces by diamond wire cutting, the last six inches (just above the PCRV. top head liner) will be removed by utilizing a mechanical breaker to break up the concrete around the perimeter of the PCRV top head liner, enabling the removal of the remaining concrete wafer in_ sections. This accident scenario assumes that radioactivity is released from the drop of a rubble transport container due to a faulty crane or operator error. 3.4.3.2 Accident Descripilgn An activation analysis performed for fort St. Vrain (Ref. 2) shows that the highest concentration of radioactivity in the PCRV concrete is in the six inch increment of the PCRV top head immediately above the top head liner as shown in . Table 3.4-3. The values in Table 3.4-3 are based on three years decay, the approximate time frame in which the dismantling work is expected to take place. The percentage contribution-of activation products within this concrete is given in Table 3.4-4. As shown, nearly 100 percent of the total activity is accounted for by the nuclides listed. L i 3.4-5 l'
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 Iron and cobalt are trace constituents in cements and aggregates. Therefore, the activation products Fe-55 and Co-60 will occur in the concrete of the PCRV, although in much smaller concentrations than in the steel rebar. However, none of the activation products , embedded in the rebar is assumed to be released in tnis postulated accident since the rebar, unlike the concrete, would remain intact upon impact. No activation _ products in the embedded rebar are assumed released. Of the entire PCRV, 5.8% by weight (2% by volume) is rebar. Based on the Activation Analysis (Appendix 11 of this plan), 32.83 Curies of Fe-55 and 1.43 Curies of Co-60 are contained within the concrete of the six inch thick concrete wafer adjacent to the PCRV top head liner, and a fraction of this activity is considered to be releasable upon impact. The remaining activity, shown in Table 3.4-3 (excluding the Co-60 and Fe-55) comprise 8% of the total activity in this 6 inch wafer (7.86 Curies). It was conservatively assumed that the remaining activity was 9.83 Curies
.(10% of total), and that 60% of this activity was Eu-154 and 40%
tritium, since tritium comprises approximately 40% of the activity and Eu-154 has the highest dose conversion factor of all the other isotopes involved. The analysis conservatively assumes that 10% of the concrete in this 6 inch thick concrete wafer is involved in the accident [approximately 7,500 lbs). Of this amount, 1% of the activity 3 the concrete is assumed to be released to the Reactor Building atmosphere. The airborne activity was calculated to be 32.8 millicuries of Fe-55,1.43 millicuries of Co-60, 3.93 millicuries of tritium, and 5.90 millicuries of Eu-154. No credit was taken for particulate filtration by the Reactor Building ventilation system. The major exposure pathway was assumed to be air inhalation by an adult standing at the EPZ (100 meters minimum) and was assumed to occur over a two hour period. The dose conversion factor used for tritium was 1.58 E-07 mrem /pCi inhaled for whole body and any organ. The dose conversion factors used for Eu-154 were 6.48 E-05 mrem /pCi for whole body dose and 7.40 E-04 mrem /pCi for bone (the highest inhalation organ dose). The dose convarsion factors used for Fe-55 were 4.93 E-07 mrem /pCi for whole body dose and 9.01 E-06 mrem /pCi for lung. The dose conversion factors used for Co-60 were 1.85 E-06 mrem /pCi for whole body dose and 7.46 E-04 mrem /pCi for lung. These / dose conversion factors were taken from NUREG-0172 (Ref. 51). The adult breathing rate was 3.47 E-04 m3/sec (Ref. 46, Section 14.12). 3.4-6 ,
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 3.4.3.3 Analysis of Effects and Consecuences The whole body and bone doses to an individual standing at a point on the EPZ 100 meters from the Reactor Building were calculated to be 4.92 mrem _ and 54.7 mrem, respectively. The whole body dose was 0.01 mrem from tritium, 4.68 mrem from Eu-154, 0.20 mrem from Fe 55 1 and 0.03 mrem from C0 60. The bone dose was 53.5 mrem from Eu-154 and 1.23 mrem from Fe-55. Co 60 and tritium did not contribute to the bone dose since neither of these have a dose conversion factor for bone. 3.4.4 Conversion ConstrRqlign_fEqsident Near PCRV Dismantlement 3.4.4.1 Identification of Casiri
- 1. Crane failure:
An evaluation was performed on the potential impact of a construction crane toppling which would impact the Reactor Building. Due to the proximity of the planned new steam generator building to the Reactor Building, it will be possible for a crane boom to strike the Reactor Building above the refueling floor level. A crane boom is relatively light and fragile. An impact with the Reactor Building is not expected to cause structural damage to the building. Additionally, LSA containers outside the Reactor Building will be protected if they are stored within the fall radius of the construction cranes. At worst, the crane boom could drape over the reactor siding. No radiological impact is expected from such an accident. This accident is bounded by the heavy load drop (Section 3.4.5) and tornado (Section 3.4,9),
- 2. Explosion / fire Due to Natural Gas Line Leak:
Fort St. Vrain will be repowered by a natural gas-fired boiler. The most severe accidents that can be postulated during decommissioning activities involve a natural gas line leak resulting in an unconfined vapor explosion or fire, or an explosion of the gas-fired boiler itself. The decommissioning and repowering schedules have been reviewed. There is over a year between completion of the removal of highly radioactive components (graphite blocks) from the PCRV and introduction of natural gas on site. In the event of a slippage in the dismantling schedule, administrative controls will be implemented to prevent charging the gas-fired boiler natural gas line on site concurrent with handling of the activated graphite blocks from the PCRV. Therefore, given the actual rchedule and 3.4-7
l PROPOSED DECOMMISSIONING PLAN o/28/91 SECTION 3 administrative controls, an explosion or fire due to a natura) gas line leak is not credible during the decommissioning process. Accidental release of activity caused by a postulated explosion of a container of flammable gas, such as those used to support decommissioning (e.g., propane or acetylene tank or bottle), was taken into consideration. Flannable liquids and gases will be administratively controlled during decommissioning and conversion to prevent use or storage of substantial quantities of flammable liquids or gas near areas containing highly activated wastes. However, even if it were postulated that an explosion did occur near radioactive waste containers, this event would not produce consequences exceeding those analyzed in this section for a heavy load drop, tornado or fire. This conclusion is based on the relatively all size of the missiles resulting from such a postulated explosion, and the relatively large amounts of activity postulated to be released in the above mentioned accidents. 3.4.5 Heavy Load Dron Accident The dismantling of the PCRV will be accomplished with the aid of three types of hoist systems. These systems include the main Reactor Building bridge crar.a, the auxiliary 17-1/2 ton hoist on the bridge crane, and three 1-1/2 ton jib cranes on the refueling floor level. The Reactor Building crane will be re-reeved to allow the 170 ton main hook to travel from the refueling floor to ground level . An elevation view of the PCRV work area is shown in Figure 3.4-3. There will be many heavy loads removed during the dismantling process. These lifts include:
- 1. Large side reflector blocks.
- 2. Large concrete sections
- 3. Steam generators.
- 4. Helium diffusers,
- 5. Concrete Core Support floor or CSF sections The accident scenarios developed for heavy load drops in nuclear power plants consider the dropping of a heavy load (e.g., fuel shipping cask) on a very large radionuclide inventory such as fuel or spent fuel (Ref. 52). In the case of Fort St. Vrain, all fuel will have been removed from the Reactor Building prior to commencement of dismantling operations. Therefore, the full spectrum of heavy load drop accidents is much less severe than in an operational nuclear power plant.
l 3.4-8 '
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 The most severe heavy load drop accident is postulated to consist of dropping the component containing the largest inventory of dispersible radioactive material. Table 3.4-5 has been compiled to show the various components and their respective radioactive inventories. Sampling will be performed prior to waste movement to determine and verify the radionuclide composition and total Curie content. Review of this table indicates that the large side reflector blocks contain the largest radioactive inventory. The use of an entire large side reflector for this accident analysL. is conservative since the predicted activity inventory exceeds the LSA Curie limit specified in 10 CFR 71, Table A-1, for Type A-2 waste containers. The drop of a heavy load onto a highly radioactive component was evaluated and determined not to represent the worst case scenario. For instance, the dropping of one of the 240 large side reflector blocks back into the PCRV might crush portions of adjacent reflector blocks. However, since all highly radioactive components are kept under water unless they are being removed, the debris and its attendant activity would remain in the water. This activity would be cleaned up in the PCRV water cleanup and clarification system, i described in Section 2.3.3.6 of this plan. Any " slosh" created by the block drop would drain back to the PCRV cavity or drain down inside the Reactor Building, eventually to the Reactor Building sump l and keyway, which have a capacity of approximately 350,000 gallons. These accident scenarios are bounded by the loss of PCRV Shielding Water accident described in Section 3.4.7. As discussed- in Section 2.3.3.10, alternatives for removal of the 270-ton concrete CSF from the PCRV include sectioning it into pieces within the PCRV, and removing - the pieces by means of the Reactor Building crane, or raising the ' entire CSF - above the PCRV with specially installed high capacity jacks. Since the activated I graphite blocks would have been removed from the PCRV prior to - removal of the CSF, and since the CSF concrete- is predicted to contain only 6 Curies of activity, a heavy load drop during this operation does not how the potential for release of significant l quantities of radioactivity. If the entire CSF is raised by high ! capacity jacks, drop of the CSF is not considered credible since , l such an accident would require multiple jack failures. i 3.4.5.1 Identification of Cause l l A heavy load drop accident is-a relatively low probability event. A L failure of the hoisting cable could cause a drop of the load. In accordance with Reference 52., the probability of this event is on 3.4-9
PROPOSED DECOMMISSIONING PLAN 6/28/91 $CCTION 3 the order of 1.0E-5 to 1.0E-6 per demand (hoist lift). The loss of the crane brakes could be due to mechanical failure, operator error, or an incorrect maintenance operation. Since the fort St. Vrain Reactor Building bridge crane does not qual i fy as a Single-Failure-Proof crane in accordance with NUREG-0554 (Ref. 53) guidelines, the loss of crane brakes is postulated as a credible failure mode. 3.4.5.2 Accident Description for this accident it is postulated that the Reactor Building bridge crane is hoisting one of the 240 large side reflector blocks. It is curm*.ly planned to section these reflector blocks into smaller pieces for packaging in LSA shipping containers. Ilowever, it is conservative to assume that a single reflector block may be transported intact in its own shipping container. After appropriate radiation surveys and removal of any surface contamination, the container with the single unsectioned side reflector block is lowered down the enlarged equipment hatch, failure of the crane is postulated at this point. This results in the side reflector block container falling approximately 100 feet to the level of the truck loading bay. The shipping container ruptures, spilling its contents on the truck loading bay floor. Administrative controls will be in place that will prevent the tractor of the tractor trailer from being in the loading bay during lowering of the container, and will ensure that all the truck loading bay doors are closed. It is conservatively assumed that one percent of the activity of a single large reflector block is dispersed from the drop. The dust is postulated to remain airborne and will escape the immediate area through the Reactor Building ventilation exhaust. Credit is taken for decontamination afforded by the Reactor Building ventilation system. The fort St. Vrain activation analysis (Ref. 2) indicates that the major contributors to the activity in these large side reflector blocks are Fe-55, tritium, and C0-60. The total activity in each of the large side reflector blocks has been calculated to be 1477 Curies. A one percent release for this scenario results in 14.77 Curies becoming airborne in the Reactor Building. Of this amount, 14.6 Curies are Fe-55, 0.06 Curies are tritium and 0.11 Curies are Co-60. These activities are based on a three year decay period. Credit is taken for a 95 percent filter efficiency for fe-55 and Co-60. Tritium is released unfiltered (Ref. 46, Section 14.12). The major exposure path was assumed to be air inhalation to an adult standing at a_p_oint on the EPZ 100 meters from the Reactor Building._ l 3.4-10
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 The dose conversion factors in mrem per picocurie inhaled were obtained from Regulatory Guide 1.109 (Ref. 54) and are as follows: Isotone Total Body Lung Tritium 1.58 E-07 1.58 E-07 Fe-55 4.93 E-07 9.01 E-06 Co-60 1.85 E-06 7.46 E-04 3.4.5.3 Analysis of Effects and Conseauences The whole body and lung doses to an individual standing at a point on the EPZ -100 meters from the Reactor Building were calculated to be 4.66 aRem and 133 mrem, respectively. The whole body dose was 0.12 mrem from tritium, 4.41 mrem from Fe-55 and 0.13 mrem from Co-60. The lung dose was less than 1 mPem from tritium, '81 mrem from Fe-55, and 52 mrem from Co-60. 3.4.6 Elrg 3.4.6,1 Identification of Cruig During decommissioning and repowering activities, there are many possible fire initiators that could result in a release of radioactive materials. .These possible fire initiators include:
- 1. Fires started from cutting torches.
- 2. Contamination control tent fire.
- 3. Fires associated with component processing activities on the refueling level.
- 4. Electrical fires.
The most likely initiator has. been determined to be a cable tray fire started from a spark during PCRV tendon cutting operations. The-fire would.be quickly extinguished by the fire watch on duty for the tendon cutting operations. The radiological consequence of this accident would be negligible since the cable trays contain. virtually no radioactivity contamination. The postulated fire accident involves a fire enveloping LSA waste containers. The greatest exposure for a fire accident to occur-is during the- approximate six month period when the highly radioactive large- side reflector blocks and side spacer blocks are being removed from the PCRV. 3.4-11 l
1 PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 Controls will be impiemented prior to the storage of the LSA containers. LSA containers will be limited to groupings with no more than the equivalent Curie content of 230 side soacer blocks. Sufficient spatial separation will be imposed to preclude fire propagation to an adjacent group of LSA containers. The packaging of these boxes and/or drums is planned to be completed inside the Reactor Building. Temporary storage or staging of these containers prior to shipment is also expected. It is assumed that interim radioactive material storage will be available for up to 15 LSA boxes and 200 drums in the former Fort St. Vrain fuel Storage Building. Fire detection capability will be installed in the LSA container storage area prior to the storage of the LSA containers. There will be no uncontrolled combustible materials in this building. The controls defined above will be implemented prior to the storage of the containers to limit the grouping of LSA packages containing combustible materials. These control s will ensure sufficient spatial separation is available to preclude fire propagation to an adjacent group of LSA containers and precludes the possibility of a fire with consequences greater than that which is analyzed. 3.4.6.2 Accident Description for the fire i.ccident it is postulated that a tractor trailer begins to transport packaged waste from the Reactor Building truck loading bay to an off-site burial ground / processing facility. The shipment is conservatively postulated to consist of 230 side spacer blocks with their boron pins removed. There are 1152 side spacer blocks to be removed during the decommissioning process. It is pustulated that an engine fire develops on the transport tractor and the fire spreads to the tractor's diesel fuel tanks. Based on work at the Waste Isolation Pilot Plant, the frequency of an unsuppressed truck fire is in the range of 1.0 E-4 to 1.0 E-5 per year (Refs. 54, 56). The tractor diesel fuel tanks may contain a combined capacity of up to 300 gallons of fuel. The fuel tanks are postulated to rupture from the heat and engulf the entire tractor trailer and the LSA containers in a diesel fuel pool fire. It is conservatively assumed that graphite side spacer blocks are enveloped by the diesel fuel fire. A fire involving 300 gallons of diesel fuel spilled onto a relatively flat surface will burn out within thirty minutes. The resultant fire temperature will be bounded by the ASTM-E119 (Ref.
- 57) standard fire curve. Most of the graphite will be exposed to 3.4-12
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 temperatures well below the fire temperatures due to insulation provided by adjacent graphite blocks and some protection afforded by the shipping containers. Under these conditions, it is conservative to assume that 50 percent of the graphite inventory on a shipping trailer is oxidized during the 30-minute fire It is assumed that all of the tritium in the oxidized fraction (50 percent of the total tritium inventory), is released. In addition to tritium release, it is assumed that 0.015 percent of tl.e bal **:0 of the radionuclide inventory is released in the form of particulates (Ref. 58). The accident is assumed to occur at ground level immediately outside of the Reactor Building truck loading bay. The radioactive inventory for the 230 graphite side spacer blocks is calculated to be 3,706 total Curtes. This total inventory consists of 3556 Curles of Fe-55, 122 Curies of tritium and 28 Curies of Co-60. Fifty percent of the tritium is assumed to be released (approximately 61 Curies). The additional release of the remaining radionuclides will be 0.534 Curies of Fe-55 and 0.0042 Curies of Co-60. The atmospheric dispersion factor, breathing rate and dose conversion factors are the same as those used for the heavy load drop analysis. 3.4.6.3 Analysis of Effects and Consecuences The whole body and lung doses to an individual standing at the EPZ were calculated to be 121 mrem and 215 mrem, respectively. As, described in Section 9 of this plan, the EPZ is a minimum of 100 meters from the Reactor Building, the fuel Storage Buildingt and the route planned for transport of activated graphite blocks between the Reactor Building and the Fuel Storage Building. Therefore, these doses are applicable at the EPZ, whether the graphite block fire is postulated to occur in the Reactor Building or its truck bay, or in the Fuel Storage Building or on the transport route anywhere between these two buildings. The whole body dose was 118 mrem from tritium, 3 mrem from Fe-55, and less than one mrem from Co-60. The lung dose was 118 mrem from tritium, 59 mrem from Fe-55, and 38 mrem from Co-60. 3.4.7 Loss of PCRV Shieldina Water Accident 3.4.7.1 Identification of Causes During a portion of the Fort St. Vrain decommissioning, the PCRV cavity will be ficoded with water. This water will be circulated and purified by the PCRV water cleanup and clarification system (Section 2.3.3.6) to gradually decrease the radioactivity in the 3.4-13
r l PROPOSED DECOMMISSIONING PLAN 6/28/91 l SECTION 3 l l water. This system S expected to be in operation during the period when the PCRV internals are being removed. This accident scenario assumes that there is a leak or rupture of the PCRV water cleanup and clarification piping resulting in a liquid release due to a mechanical impact or a mechanical failure of a weld or flange. 3.4.7.2 Accident Descrintion This accident scenario assumes that a mechanical failure of the PCRV water cleanup and clarification system piping to the PCRV cavity l i occurs, resulting in a pipe rupture. Tritiated water with dissolved - l cesiuin, iron and cobalt would be spilled into the Reactor Building sump and keyway. Assuming the worst case (complete emptying of the PCRV), calculations indicate that 423,500 gallons could fill the i Reactor Building sump /kuway, and flood the basement floor to a height of two feet. This water would be 49 feet below grade and would be contained by the Reactor Building sump / keyway and walls. No credit is taken for the Reactor Bui; ding ventilation system for this accident scenario. Since the non gaseous activities will be retained in the spilled water, tritium (released through evaporation) is the only significant activity available. This will be evaporated from, the surface area of the spilled water in the Reactor Building basement. The PCRV liquid release will not seep through the sump concrete seams as the water table is well above the 49 foot below grade level. To date, no known in-leakage of ground water has been observed into the Reactor Building sump. The Reactor Building is approximately 120 feet long and 76 feet wide, which conservatively provides (neglecting c cipment) a surface area for the spilled water of 9120 square feet (848 square meters). From Wcstinghouse Report WCAP 11002 (Ref. 59), the best fit i evaporation rate at 70 percent relative. humidity and wind speed of 1 l m/sec is 0.046 g/m2.sec or 0.046 cc/m2.sec (assuming 1 gram - Ice of l water). It is precicted that tritium levels in the PCRV water will be less than 1000 Curies. However, for this analysis, it is conservatively assumed that the theoretical maximum . amount of tritium is transferred to the PCRV shielding water from the graphite b1<>r.ks, which is approximately 1 E+5 Curies. Therefore, the tritium l concentration in the spilled water is calculated to be 62.4 uCi/cc. l 3.4-14 \
-- - -. - - . - = - - . - _ . .
- . _ . - ~
l l l PROPOSED DEC0iVilS$10hlNG PLAN 6/28/91 SECTION 3 3.4.8.2 Accident Description loss of power would result in the loss of plant ventilation (IIVAC) systems, lighting, plant water systems, and demolition power. Decommissioning activities would cease until power is restored. Loss of power to the PCRV water cleanup and clarification pumps will not result in a radioactive release since the flow of bleed water to the evaporation ponds will be stopped (see Section 3.3.2.2 for a do.cription of this process). While loss of ventilation will force personnel from radiological control areas, no off site consequences are anticipated. The postulated accident scenario is the loss of power to the HVAC while a large side reflector block has been removed from the PCRV for cutting. These graphite blocks will be grappled and hoisted by a jib crane to a refueling floor work station. At a work station a block will be cut into sections in preparation for packaging into LSA containers. The loss of power is assumed to occur af ter the cutting / cleaving operation. It is assumed that these processing operations (kerfing debris) release 1.5 percent of the total activity of a single large side reflector block, it is conservatively judged that the combination of radiological controls in place at the w rk station (e.g., confinement through tenting) and the confinement function provided by the Reactor Building itself will result in retention of 99 percent of the re 55 and Co-60 kerfing debris in the Reactor Building. It is assumed that one percent of the fe 55 and Co 60 in the kerfing debris and 100 percent of the tritium in the kerfing debris are released at ground icvel from the Reactor Building. No credit is taken for the Reactor Building ventilation system. The intal activity in each of the large side reflector _ blocks has been calculated to be 1477 Curies. A release of 1.5 percent of the , radioactive material is assumed from the kerfing debris in each l block. Of that amount, one percent of the re-55 and Co 60, and 100 percent of the tritium is released, resulting in a total of 0.31 l- Curies released to the environment. This total release consists of l 0.219 Curies of re 55, 0,091 Curies of tritium and 0.0017 Curies of L Cc 60. These activities are based on a three year decay period. l The major exposure path was assumed to be air inhalation to an adult standing at aj oint on the EPZ 100 meters from the Reactor Building. The atmospheric dispersion factor, breathing rate, and dose conversion factors are the same as those used previously. 3.4-16
6/28/91 PROPOSED DECOWilSS10NING PLAN i SECTION 3 3.4.8.3 Analysis of Effects and Consecuences: ' The whole body and lung doses to an individual standing et a point on the EPZ 100 meters from the Reactor Building. were calculated to be 1,54 mRom and 40.0 mrem, respectively. The whole body dose was 0.18 mrem from tritium, 1,32 mrem from Fe-55, and 0.04 mrem from Co-60. The lung dose was 0.2 mrem from tritium, 24.2 mrem from Fe-55 and 15.6 mrem from Co 60. 3.4.9 Natural Diststers For the effects of natural disasters, the following external initiating events were considered:
- 1. . Earthquake The Reactor Building is designed to withstand the Design Basis Earthquake of- 0.10 g horizontal ground acceleration at the site without unsafe damage or failure to function. For decommissioning, it is required that the Reactor Building continue to perform its confinement function following a seismic event. The seismic qualification of the Reactor Building will be maintained during decoaimissioning. No other new or existing systems or equipment are required to function during or following an earthquake.
The most severe event which could result from a large earthquake is considered to be a drop of a radioactive waste container holding a highly activated graphite block (see heavy load drop accident). However, the simultaneous occurrence of an ecrthquake and the , hoisting of a heavy load is not considered credible (a probability i _ of less than 1 E 6 per year, from Ref. 52). The consequences of this simultaneous earthquake and heavy load drop scenario were not analyzed due to the low probability of such an event.
- 2. Tornado and Wind Effects from Reference 46, Section 14.1.2, the_ basic design wind velocity for the- plant is 100 mph. The equipment and structures exposed to l
wind load are designed to support design wind load combined with functional loads within the specified allowable stresses. The tornado danger at the plant site is extremely remote. However, the Reactor Building was designed to withstand wind loadings developed by a tornado of 202 mph (total horizontal wind velocity) without exceeding yield stresses in the basic building structure. The Reactor Duilding was also designed to withstand a maximum 3.4 17
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 i i tornado of 300 mph (total horizontal wind velocity) acting on the full area of all structures and a drop in atmospheric pressure of 3 psi within a period of 3 seconds, without exceeding ultimate stress levels in the main structural members. Above the 202 mph wind speed, the siding on the Turbine and Reactor Buildings above the turbine deck and refueling floor levels may be carried away, but the basic ba 1 ding structure will not collapse.
- 3. Floods from Reference 46, Section 14.1.3, the plant site is protected from excessive runoff and flood by design of the yard drainage system.
_ Grado level is approximately 17 feet above the highest observed flood level, and from 10 to 13 feet above the maximum probable flood level. The walls of the structures extending below grade level are watertight, and buoyancy effects were taken into account in their construction. Therefore, there will be no further consideration of accidents due to flooding during decommissioning activities. >
- 4. Range Fire The Fort St. Vrain site is located in an area of Weld county devoted ,
to agriculture. The site itsel f is mostly surrounded by corn l fields. Within the plant exclusion area is a fire buffer area
-consisting of maintained grass and ornamental land::caping. A 20 foot wide concrete pad rings the site. Therefore, a brush or range fire is not a credible accident dtaing decommissioning activities.
3.4.9.1 Identification of Cwin The risks from a tornado at Fort St. Vrain during decommissioning are quite low for two reasons, first, the probability that a tornado will strike the site is diminishing 1y small. Second, the plant specific. vulnerability to a tornado and its consequences are also small. Unlike an operating nuclear power plant with active safety systems 'to contain large quantities of radioactive materials ' at high energy levels, all spent fuel will be removed from fort St. Vrain _and _the PCRV ' will essentia11y'_be a passive _ container of radioactive material. _Possible loss of power caused by a-tornado.is specifically analyzed in Section 3.4.8. The Reactor Building roof _ and siding above the refueling floor are designed to withstand a tornado with a wind speed up to '202 mph. The probability of_ experiencing a tornado with wind speeds above 202 aph during decommissioning is extremely low based upon information 3.4-18
i 6/28/91 PROPOSED DEC0WilSS10NING PLAN SECTION 3 and methodology provided in the draft Individual Plant Examination
- of External Events (IPEEE), NUREG-1407 (Ref. 60).
l Based on the work of Abbey and Fujita (Ref. 61), the continental United States was broken down into 20 distinct tornado hazard regions. These regions were generalized into 4 broad areas shown in Figure 3.4-4, ranging from a highest risk in region A to the lowest risk it. Region D. The fort St. Vrain site is classified into Region C. Reference 62 is used to establish the occurrence rate for different classifications of tornadoes. The National Severe Storms Forecast Center (NSSFC) national database for the years 1950 - 1978 was used as the basis for the occurrence rate analysis. The NSSFC data are categorized .by Fujita intensity scales (F-scales). To predict the probability that a tornado with maximum windspeed will strike a nuclear power plant requires adjusting the F-scales for: tornado reporting trends, F-scale classification errors, path length intensity variation, and occurrence rates and windspeed relationships adjusted for intensity variation. The adjusted, or updated, tornado scales are denoted by "F'". Tornado wind velocities for the F- and F'- scales are compared as follows: Maximum Windspeed Maximum Windspeed F-Scale Interval (moh) F'-Scale Interval (mnh) F0 40 - 72 F'O 40 - 73 F1 73 - 112 F'l 73 - 103 ' F2 113 - 157 F'2 103 - 135 F3 158 - 206 F'3 135 - 168 F4 207 - 260 F'4 168 - 209 F5 261 - 318 F'S 209 - 277 The following evaluation demonstrates the low probability of occurrence of a tornado with wind velocity exceeding 202 mph at Fort St. Vrain, by comparing the frequency of occurrence of tornadoes in Weld County with the NSSFC data. The occurrence rate of a F'4 tornado is 3.4 E-6/ square mile / year (of. 62). According to the National Weather Bureau's historical data for Weld County from 1950 through 1987 - there was only one tornado in- the F3 range. That single F3 tornado is the only tornado in the vicinity of Fort St. Vrain of the 256 tornadoes recorded by NSSFC for all of Region C that had estimated windspeeds greater than 158 mph (Ref. 62). Based on this sample from the population, it can be inferred that the probability of a tornado at Fort St. Vrain in the F3 range is much less than 3.4 E-6/ square mile / year. 3.4-19
. .~ .-. . - - -. - . -. ..-_ -
PROPOSED DECOMMIS$10NING PLAN 6/28/91 SECTION 3 1he occurrence rate for a f'5 tornado in Region C is 3.5 E 7/ square mile / year (Ref. 62). The National Weather Bureau's Weld County data show no tornado occurrence with intensity of f4 or greater. Thus, the 56 f4 and nine FS tornadoes recorded by NSSFC all occurred outside the fort St. Vrain area, from this data, it can be concluded that the probability of occurrence of an f4 or greater tornado is less than 3.5 E-7/ mil /yr. According to the draft IPEEE " Plants Designed Against NRC Current Criteria", these events pose no significant threat of a severe accident because the current design criteria for wind are dominated by tornadoes having a frequency of exceedance of about 1 E-7. lhe following section contains a specific accident analysis for a postulated tornado with winds less than 202 mph. 3.4.9.2 Accident Description Temporary storage or staging of ra& active waste containers prior to shipment is expected. It is assumed that interim radioactive material storage will be available for 15 LSA boxes and 200 drums in the fort St. Vrain fuel Storage Building (Section 2.2.1.3). Calculations demonstrate that neither forces generated by 202 mph wind loading, nor the impact frem the tornado-driven design basis missile, sill result in breach of the walls or roof of this building. In this scenario, it is assumed that a 202 mph tornado strikes the fort St. Vrain site. At this lower wind level, the walls of the Reactor Building enclosing the PCRV will remain intact. The tornado driven design basis missile is a 12 foot x 12 inch x 4 inch thick fir plank, weighing 105 pounds, which impacts and penetrates the Reactor Building above the refueling floor level. It is-assumed that this missile strikes and ruptures a container with 46 graphite side spacer blocks, it is conservatively assumed that One cercent of the activity in the container is dispersed and released to the environment. No filtration credit is assumed. The total radioactivity inventory for the 46 side spacer blocks is approximately 741 Curies. This total inventory is comprised of 711 Curies of- fe-55, 24.4 Curies of tritium and about 5.5 Curies af Co-60. Assuming a one percent release results in 7.41 Curies released to the environment. These activities are based on a three year decay period. The major exposure path was assumed to be air 3.4-20 t
, .- - .. - . , . . . ~
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 inhalation to an adult standing at a_ point _on the FPZ 100 meters from the Reactor Building. l The atmospheric dispersion factor used was an annual average fort St. Vrain dispersion, factor of 4.59 E-4 sec/m3, calculated for 100 meters (the minimum distance to the EPZ), and assuming ground level release. This is considered very conservative, since during a tornado or in the wake of a tornado the atmospheric dispersion factors would be much more favorable. The dose conversion factors i and breathing rate used are the same as those stated previously. l 3.4.9.3 htalysis of Effects and Conseauences: H 1he whole body and lung doses to an individual standing at a_ point .
-on the EPZ 100 meters from the Reactor Building were calculated to be 0.58 mrem and 16.8 mrem, respectively. The whole body dose was i 0.006 mRom from tritium, 0.558 mrem from Fe 55, and 0.016 mrem from _i C0 60. The -lung dose was less than 0.01 mrem from tritium, 10.2 mrem fro. Fe-55 and 6.6 mrem from C0 60.
3.4.10 Summary The results of the preceding accident scenarios, postulated for fort St. Vrain decommissioning activities, indicate that the radiation exposures to the general public will be very low. These evaluations have determined that, in all cases, the radiological consequences at the EPZ (100 meters minimum) are well within the 10 CFR 100 guidelines of 25 Rem whole body dose and 300 Rem to any specific organ. These doses are also a small fraction of the one Rem whole body dose and five Rem to any specific organ guidelines cited in the EPA Protective Action Guidelines (Ref. 48). These scenarios are considered to have a low probability of occurrence and their radiological consequences bound other less severe accidents scenarios. Therefore, it is concluded that the fort St. Vrain decommissioning activities do not pose any undue risk to the health and safety of the general public, i 3.4-21
_,,m, , _ , , PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3 i l l INTENTIONALLY LEFT BLANK 1 3.4-22
-- ..'~, -e,,. -_-._s,-,. . . - . - . .. ..m s. ,, - - . - - e,,. . , _-.. ,m.w,,.,...,,.www.- ,.-,-w_,--, -v ,- , ,,.e,--,_--,-m- -
6/28/91- PROPOSED DECOMMISSIONING PLAN SECTION 3 TABLE 3.4-1
SUMMARY
OF ACCIDENT SCENARIOS KC11QI DELCRIPTION Dropping of Contaminated Rubble from PCRV top Concrete Rubble head concrete is dropped during processing Conversion Construction ifatural gas explosion / Near PCRV Dismantlement crane falling. Heavy Locd Drop Container drop to loading bay. Fire Truck diesel fuel pool fire. Loss of PCRV Shielding Water Pipe rupture in the PCRV water cleanup / clarification system. Loss of Power Release of graphite cutting debris from refueling floor work station. Natural Disasters Tornado generated missile striking LSA waste l container. l l 1
I PROPOSED DECOMMISSIONING PLAN 6/28/91 SEC110N 3 TABLE 3.4 2 DOSES TO AN INDIVIDVAL AT Tile EPZ (100 METER MINIMUM 1 ~ ! DUE TO POSTULATED ACCIDENTS 2-HOURDOSE(MREM) ACCIDENT WHOLE BODY DOSE ORGAN DOSE Dropping of Concrete Rubble 4.92 54.7 (bone) Heavy Load Drop 4.66 133 (lung) Fire 121 215 (lung)
-5 Loss of PCRV Shielding Water 34.8 34.8 (lung)
Loss ~of Power 1.54 40.0 (lung) Natural Disaster (Tornado) 0.58 16.8 (lung) t= l-l
. . - . , _ . . - - . . . _ _ . . . . . . . _ . . . _ , . _ . - _ , _ ~ . _ . - - , _ _ . , _ _ . _ , , _ - . , , , . , , - _ - - - _ _ - . -
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 101 g
~
Anticipated 10 CFR 50 100 r Operational
- Appendix I
" Occurrences W jg-1 _ Region / $~ [ Sheltering Requirement .. r-----e-~~~--~~-- /
J--- - 2.5 x 10 2 8 10-2 7 iC : Design w Basis, 2 10-3 r Region b -
~
10 CFR 100
$104 r-----------------< ----
Acute 1.0 x 10-4 if : Fatality g . Emergency Safet) 10-8 r Planmng Goal E : Basis j - Region t------------------<------- 5.0 x 10-7
~
10-7 r 104>iiliiiili>>d *iid iiiiliiiilii>>l i >>li**J i>>J >>>< 10-* 10-5 10d 10-3 10-2 10-1 100 3o1 10 103 108 Mean Whole Body Gamma Dose at EPZ (Rem) s figure 3.4-1 Whole Body Exposure Guidelines at the Emergency Planning Zone (EPZ)
-- -, ,.,--n-, . , . , , , - . . , - - - - ,, ,-,m- - - - , - r ,
PROPOSED DECOMMISSIONING PLAN 0/28/91 SECTION 3 l l l f f 1000 215 121
- E Whole Body 54.7 dMF ma" z' N organ 100 8EIF 34.8 34.8 40.0 muura mer meer f #
16.8 F, '
'"F
.- c -
4.9 4.7 " r-- E 10 ' A EIB g . . m 1.5 o l O , .' mim -
.6 mas Rutuo en Load Dro Fire PCRV Breach oss of Power Tornado Postulated Accident l
l l l f figure 3.4-2 Dose Consequences from Fort St. Vrain Postulated Accidents
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3
- 12. NRC .
Regulatory _ _ Guide 8_. l.01 " Operating Philosophy _for Maintaining _ Occupational Radiation Exposures as low as is Reasonably Achievable", Rev.__1R. May 1977.
- 13. NRC Regulatory Guide 8.15 " Acceptable Programs for R espiratory Protection", October 1976.
J4. NUREG 004L " Manual of Respiratory Protection Against Airborne Radioactivity Materials", October 1976. J5. NRC Regulatory Guide 1.8 "quali fication and Training of Personnel for Nuclear Power Plants"2 Rev. 2,_ April 1987.
- 16. ANS/ ANSI 3.1 " Selection, Training and _ Qualification of Personnel for Nuclear Power Plants"
- 17. NRC Regulatory Guide 8.27, " Radiation Protection Training for Personnel At Light Water Cooled Nuclear Power Plants", Marct!
1981
- 18. NRC Regulatory Guide 8.13 " Instruction Concerning - Pre-natal Radiation Exposure", Rev. 2,_ December 1987.,
- 19. HRC Regulatory Guide 8.29, " Instruction Concerning Risks frc..i Occupational Radiation _E pos..re", July 1981.
- 20. NRC IE Bulletin 79-19 " Packaging of low Level Radioactive Waste for Transport and Burial", August 10, 1979.
, 21. NRC Regulatory Guide 8.4, " Direct-Reading and Indirect-Reading L Pocket Dosirreters", February 1973.
- 22. NRC Regulatory Guide' 8.14, " Personnel Neutron Dosimeters",
Rev. 1, August 1977.
- 23. ANSI N13.ll, " Criteria for Testing Personnel", 1983.
"A J ylication of Bioassay for ~
- 24. NRC Regulatory Guide 8.26, Fission and Activation Products", September 1980.
- 25. NRC Regulatory Guide 8.9, " Acceptable Concepts, Model s, Equations and Assumptions for a Bloassay Program". September 1973
- 26. NRC Regulatory Guide 8.32 " Criteria for Establishing a Tritium Bioassay Program", July 1988 3.5-2 )
l l 1
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 3 ,
- 27. ANSI /CGA G-7.1, " Commodity Sgecification for Air", 1989 28 NRC Circular 81072 " Control of Radioactively Contaminated Materials," May 14,_1981.
l
- 29. NRC IE Notice _85 92, " Surveys of Wastes Before Disposal from i Nuclear Reactor facilities", December 2, 1985. I
- 30. NRC Regulatory Guide 1.86, " Termination of___ Operating Licenses for Nuclear Reactors", June _1_974
- 31. Colorado Department of Health,- Radiation Controls Division,
" Soil Contamination Guidance Policy "-Rev. 1, Dated 10 31 83 1
- 32. NRC Regulatory Guide 8.25 " Calibration and Error Limits of Air I Sampling -instruments for Total Volume of Air Sampled" August 1980 3
- 33. NRC Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)", Rev. 2, February 19782
- 34. ANSI N13.1-1969, "American National Standard _ Guide to Sampling Airborne Radioactive Material in Nuclear Facilities", 1969.
- 35. ANSI N42.14 1978, " Calibration and Usage of Germanium Detectors for Measurement , of Gamma Ray Emission of R_adionuclides", 1978.
- 36. ANSI N42.3-1969, "American National Standard and IEEE Standard Test Procedure for Geiger Mueller Counters",1961
- 37. ANSI N320-1979, " Performance Specifications for Reactor
-Emergency Radiological Monitoring Instrumentation", 1978.
- 38. ANSI N323 1978, " Radiation Protection Instrumentation Test and-Calibration",-1978.
- 39. ANSI /IEEE Std 325 1986, "lEEE Standard Test Procedures for Germanium Gaana-Ray Detectors",1986.
- 40. .The As Modified Three Party Agreement, Document Number 34426, Daced July 1, 1965.
- 41. DOE letter, K.R. Hastings (DOE) to R. Husted (PSC), dated May l 11, 1588 (G-63163);
Subject:
" Modification to Contract at (04 3)-633: DOE Commitment to Receive FSV Spent Fuel".
3.5-3 != 1 _ . . _ _ . _ , _ __ _ _ . - -
l PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 3
- 42. PSC Letter, Crawford to Roberts (NRC), dated June 22, 1990 (P-90173);
Subject:
" fort St. Vrain ISFSI Licente Application".
_4 3 . NRC Regulatory Guide 1.52 " Design, Testing, and Maintenance Criteria for Post-Accident Engineered Safety feature Atmosphere Cleanup system Air Filtration and Adsorptio,1 Units of Light Water Cooled Nuclear Power Plants", Rev. 2, March, 1978.
- 44. NRC Generic Letter 81-38 " Storage of Low. level Radioactive Wastes at Power Reactor Sites", November 10, 1981.
4_5 . .NUREG 0800 " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants", July 1981.
- 46. Fort St. Vrain Updated final Safety Analysis Report, Rev. 8, Public Service Company of Colorado.
- 47. Nuclear Safety, Volume 31, No. 2, April June 1990.
- 48. " Manual of Protective Action Guides and Protective Actions for Nuclear Incidents", EPA 520/1 75-001 A, U.S. Environmental Protection Agency, January 1990.
- 49. General Atomics Internal Correspondence, M.B. Richards to F.
C. Dahms, " Tritium Source Terms in fort St. Vrain Permanent Side Reflector and Spacer Block Graphite", dated October 10, 1990; attachment to General Atomics Letter GP-3487, dated October 11, 1990.
- 50. NRC Regulatory Guide 1.145, " Atmospheric Dispersion-Models for Potential Accident Consequence Assessments at Nuclear Power P1 ants," November 1982.
- 51. NUREG-0172 " Age-Specific Radiation Dose Commitment factors for a One Year Chronic Intake", October 1977.
- 52. 'NUREG-0612 " Control of Heavy Loads at Nuclear- Power Plants",
July 1980. L '
- 53. NUREG 0554 " Single-failure-Proof Cranes for Nuclear Power P1 ants",.May 1979.
L L l l l 3.5-4
l l I 6/E8/91 PROPOSED DECOMMISSIONING PLAN i SECTION 3
- 54. NRC Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CfR Part 50," Appendix 1, Revision 1, October 1977.
- 55. " Waste Isolation Pilot Plant Integrated Risk Assessment," DOE )
89-10, June 1990. j
- 56. *5pecial Analysis: Heavy Truck fires, 1982-1986 " National fire Protection Association fire Analysis and Research i Division.
- 57. ASTM E119-80 " Standard Test Mathods for fire Tests of Duilding Construction and Materials". )
- 58. Mishma, J. and Schwendiman; " fractional Airborne Release of ;
Uranium During the Burning of Contaminated Waste," BNWL-1730, April 1973.
- 59. Westinghouse WCAP-11002, " Evaluation of Steam Generator j Overfill Due to a Steam Generator Tube Rupture Accident,"
february 1986 (proprietary).
- 60. NUREG 1407 "Procedu.al and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities" (Draft), July 1990.-
Abbey, R.f., and fujita, T.T.; 'Regionalization of the Tornado 61. Hazard, Tenth Conference on Severe local Storms," American Meteorological Society, October 1977.
- 62. " Tornado Missile Simulation and Design Methodology, Volume 2:
Model Verification and Data Base Updates", EPRI NP-2005, Volumo 2, Project 616-2, final Rtport, August, 1981. l a l l 3.5 5
PROPOSED DECOMMISSIONING l>lAN 6/28/91 SECTION 3 i l l l 1 l-INTENil0NALLY LEFT BLANK i l t I ( ' l I 3.5-6 - l.
i 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 4 SECTION 4 FINAL RADIATION SURVEY PLAN
4.1 INTRODUCTION
The purpose of the finsi radiation survey will be to demonstrate the effectiveness of the decommissioning and to provide documentatdon that contaminated materials, structures, areas and components have been successfully removed / decontaminated to acceptable levels to permit release for unrestricted use. The final radiation survey to release the fort St. Vrain site, facilities and installed equipnent for unrestricted use will be performed following the completica of the decontamination and dismantlement activities. Materials and equipment determined to be free of radicar.tive contamination will be unconditionally released on an on going basis. All radiological surveys will be conducted in accordance with approved procedules using techniques that determine the effectiveness of a particular dismantlement and/or decontamination effort. _These surveys will indicate when no further decontarination is needed and indicate that the equipment, area or structure has been prepared for unrestricted release. This section describes the proposed methodology and criteria that will 's used in performing the final surveys. This includes defini ion of residual radioactivity limits (includir,g background evaluation), radiation survey methods, material release criteria and site release criteria. 4.2 flNAL RELEASE CRITERIA The release of the site, facilities and materials remaining on site will be based on proper application of r
- face centamination, l soil / water concentrations and exposure rate rt aase criteria. While l each criterion introduced below has been derived in a 'nanner unique l
to its. radiological category (concentrations, contamination or exposure), the basis for each criterion is the same as the objective of the decommissioning _ effort itself, to insure tSat the final disposition of Fort St. Vrain will not pose a significant threat to the general health and safety of the public and can be released for unrestricted use.
- 1. Limits for loose and Fixed Surface Contamiration: Criteria to allow release for unrestricted use for both loose and fixed surface contamination have been established in NRC Regulatory Guide 1.86 " Termination of Operating 41
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 4 l ticenses for Nuclear Reactors" (Ref.1). These limits for acceptable surface contamination levels were established in 1974 and are currently accepted as practical criteria when considered in light of the maximum sensitivity of commercially available portable radiation survey equipment. All final surveys for surface contamination on materials, equipment and structures at fort St. Vrain to be released for unrestricted use shall be based on this criteria.
- 2. Limits for Direct Exposure: While re formal criteria exist that establish an acceptable level of direct exposure, the NRC has provided interim guidance which directs licensees to use a limit of 5 microR/hr above background (at one meter) for reactor generated gamma emitting isotopes as a limiting icvel for direct exposure from " residual" l radioactivity (Ref. 2, 3, 4). This recommended limit of 5 microR/hr is also consistent with statements within NUREG 0586, " final Generic Environmental Impact Statement on Decomissioning of Nuclear facilities" (Ref. 5).
While this criteria will be of use for certain unique evaluations, it is expected that the NRC Regulatory Guide 1.86 criteria will be the most restrictive for the majority of material, equipment and facility release for unrestricted use,
- 3. timits for Total Concentrations in Soil and Water: In february 1990, the NRC released NUREG/CR-5512 " Residual Radioactive Contamination- from Decommissioning" (Ref. 6) for comment. In this report, a generic ' pathway model is used to derive the potential total effective dose equivalent (TEDE) to an- individual in- a given population group from unit radionuclide concentrations of residual contamination. In consideration of this document, the effective criteria for the total concentrations of radioactive materials above- background in soil and water will be based upon those established in NUREG/CR-5512. The use of these concentrations (or methodology used to obtain these concentraticns) will ensure an average total effective dose equivalent of less than 10 mrem /yr to an individual in a given population group'.
- 4. Limits for Unrestricted Release of Decontaminated items:
Equipment and materials- found to be free of radioactive t contamination as described in NRC Circular 81-07 (Ref. 7), ! " Control of Radioactively Contaminated Haterials" and NRC 4-2 l -
6/28/91 PROPOSED DECOMMISS10NillG PLAll SECTION 4 IEN 85-92 (Ref. 8), " Surveys of Wastes before Disposal from Nuclear Reactor facilities" will be unconditionally released. Equip a t and material that are found to be contaminated and cannot be decontaminated will be handled as radioactive waste. These contaminated materials will be packaged and shipped in the most cost effective way to a radwaste volume reduction facility or to a burial site for final disposition. 4.3 SURVEY METHODOLOGY The survey methodology provides the framework for the design of survey techniques and procedures to accomplish the objective of demonstrating that the fort St. Vrain site meets all applicable radiological criteria prior to its release for unrestricted use. The final radiation survey will be performed after dismantlement / decontamination have been completed and will be based on categorizing portions of the plant and site into areas where a high, medium or low probability will exist of finding measurable amounts of residual radioactivity. Areas with a high probability of residual activity will include such areas as the PCRV, fuel storage facility, liSF and radwaste compacting building. Surfaces in these areas will be systematically surveyed. Statistical methods, described in fiUREC/CR-2082, "Honitoring for the Compliance with Decommissioning Termination Survey Criteria" (Ref. 9), will be used to determine systematic, stratification or random survey techniques. The initial site characterization survey (See Section 3.1.5), decontamination surveys during decommissioning, and routine health physics surveys will be used to determine those plan areas in which there will be a high, medium, or low probability of finding residual radioactivity. Areas with a medium probability of residual activity will typically include balance of plant areas (where contaminated equipment had been removed), ventilation systems, and contaminated equipment storage. A stratification survey technique will be used in these areas. floors and walls up to two meters above the floor will be surveyed systematically. Surveys of the ceiling and remaining wall surfaces will employ random survey techniques, since the entire area will have been previously decontaminated and will have a low probability of being contaminated. 4-3
1 PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 4 Areas with a low probability of residual activity will include such areas as the ground and structures outside the reactor building, roof, and walls of the reactor building above the operating floor, and other working areas inside the plant. Surveys of these areas will employ random survey techniques. 4.3.1 Eteliminary Survey The elements of a preliminary survey program have been incorporated into the site characterization study (See Section 3.1.S). Data obtained from the site characterization study, reviews of historical radiological data and on going radiological surveys will be utilized to develop the final radiation survey plan. This approach will provide sufficient data to preclude the necessity of developing a separate preliminary survey. This approach will provide the level of detail necessary to determine the appropriate surveys, analyses and logical division of the site into separate survey units or grids (See Section 4.3.3). 4.3.2 Backaround Determination Bade'ound levels of radiation will be determined principally by takt..g radiological measurements at an area (or areas) re .ote enough to be beyond any detectable influence of the plant, but close enough to the plant to be representative. Background measurements will include both " instrument background" and naturally occurring radioactive materials including enhanced background radiation levels due to fallout. Efforts will be made to find a site and structure that meet the above conditions and approximate the physical characteristics of Fort St. Vrain. The sampling scheme (sample locations, number of samples, etc.) will be based on guidance from NUREG/CR 2082 (Ref. 9). The sampling methodology to be used when determining the independent radiological background will be based upon collecting multiple samples of soil and direct instrument readings. Since background levels may vary from point to point, each type of background sampling will be statistically analyzed to determine if a single numerical representation of the background type will be valid. For example, it is expected that background gamma dose rates at all elevations inside buildings will be statistically equivalent to the background gamma dose rates at ground level in the buildings. If statistical differences in background levels are found at some elevation (or area), different background levels will be assigned to those areas. Statistical analyses of data, including treatment of 4-4
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 4 anomalies, will be performed based on the guidance of NUREG/CR 2082 (Ref. 9). Radiological background types that will be evaluated include:
- 1. Direct surface beta, gamma and altha contamination.
- 2. Direct gamma exposure rate readings on contact with the surface. "
- 3. Direct gamma exposure rate readings at a fixed distance (one meter) from the surface.
- 4. Surface soil contamination. .
- 5. Sub-surface soil t.ontamination.
As part of the background characterization study, the minimum detectable activity (MDA) of the instrumentation will also be evaluated to ensure that the instrumentation is sensitive enough to respond to levels as specified in the final release criteria. NRC Circular 81-07, " Control of Radioactively Contaminated Haterials" (Ref. 7) and NRC IEN 85-92, " Surveys of Wastes before Disposal from Nuclear Reactor facilities" (Ref. 8) provide guidance for the determination of the MDA of the survey instrumentation. The effects of concrete and other shielding (building geometry) within the fort St. Vrain facility that lower the background dose rate relative to the remote area readings will also be addressed and evaluated. Water and soil samples will also be taken at the remote location (s) for background evaluation. Results of this study will be documented and the formal report and final interpretation will be the basis for background baselines. 4.3.3 Grid Survey Technion To assure that all areas of a surface are adequately surveyed, a rectangular or other appropriate geometric grid will be superimposed on. all surfaces being surveyed. The grids may be physically marked on the surfaces or, as a minimum, the grid corners will be labeled. The primary purpose of the grid- is to aid in repeatability of measurements in the event that further evaluation of data is necessary. The grid dimensions will vary from one to three meters on a side per
. grid for indoor areas and certain outdoor areas (such as rooftops) ,
and from three to fif ty meters on a side per grid for soil and equipment lay-down areas. The soil grid will be laid out using stakes as markers to define grid patterns. Radiation survey maps i 4-5 l _. - -~ . _ _ _ . _ _ _ _ -
l PROPOSED DECOMMISSIONING Pl.AN 6/28/91 l SECTION 4 will be developed and included in the procedures for radiological survey. Dett.iled development of grid survey techniques and procedures will be based upon guidance from NUREG/CR-2082 (Ref. 9).
- 4. 3. 4. Special Surveys Final survey plans are typically based on the assumption that the majority of the original equipment will be removed as part of the ;
decommissioning project. Decommissioning of Fort St. Vrain is unique in that the final survey program will include the release for unrestricted use large amounts of equipment and materials that will be re used following the conversion of the facility or left in place. Electrical conduits, pipes, drains, equipment, and associated support equipment will_ require different survey methodology than a formal grid survey methodology. For these types of surveys, special techniques will be developed. Due to the large amount of equipment and materials planned to remain in place in the Reactor Building, special surveys will not encompass 100% of all piping, conduits or systems on site. For most secondary systems, this approach will be warranted due to the operational history of the facility and past operational surveys. In general, l the number and type of measurements will be based on the
- accessibility and the probability of contamination for a particular area, system, or equipment. It is expected that the site characterization survey and on-going surveys during the decommissioning will define the extent of special surveys, Techniques and procedures will be developed to ensure proper surveys of all equipment and material types (motors, vessels, piping, etc.).
Equipment or material found to be above the release criteria levels specified in Section 4.2 will be decontaminated or dismantled for disposal.
, 4.4 INSTRUMENTATION Instrumentation to be used for the final site sarvey will be of such types and ranges to ensure that measurements can be performed within the final release criteria limits.
Portable field instruments will be chosen for their sensitivity, durability, ease of use, accuracy, and portability. This class of I instruments will typically include: L 4-6
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 4
- 1. Ratemeters with thin window GM tube detectors (" pancake" Type) sensitive to gross beta radiation.
- 2. Ratemeters with scintillation or air proportional detectors sensitive to gross alpha radiation.
- 3. MicroR meters sensitive to gross gamma radiation.
- 4. Ratemeters with scintillation detectors sensitive to gross gamma radiation.
- 5. Portable multichannel analyzer with HPGo or Geli detector for field gamma spectral analysis.
- 6. Portable scaler (s) with detectors sensitive to alpha, beta and or ganea radiation.
Laboratory instruments will also be chosen for their sensitivity, durability, case of use, accuracy, and stability. This class of instruments will typically include:
- 1. Multichannel analyzer with HPGe or Geli detector (s) for gamma spectral analysis,
- 2. Liquid scintillation counter with adjustable window (s).
- 3. Scaler (s) with scintillation or gas flow proportional detector sensitive to gross alpha radiation.
- 4. Scaler (s) with GM or gas flow proportional detector sensitive to beta-gamma radiation.
Instruments will be calibrated, maintained and repaired in accordance with procedural requirements. Calibration sources to be used for calibration of both field and laboratory instrumentation will be traceable to National Institute of Standards and Technology (NIST) or equivalent standards. Procedural guidance will also be provided for a quality assurance and control program for all instruments used as part of the final survey plan. 4.5 DOCUMENTATION Survey data will be presented in a manner that will allow the final radiological condition of the site to be completely and accurately depicted. This will allow parties to ascertain the radiological condition of the site without further analysis of the data. Clear and accurate documentation will be provided to ensure acceptable agreement between the final survey, independent verification survey (Section 4.7), and the NRC confirmatory survey. 4.5.1 Survey Documentation Procedures will be. developed to provide guidance in the documentation of measurements and analytical results. Survey maps 1 4-7 l
PROPOSED DECOMMISSIONING PLAN 6/28/91 SEC110N 4 will be used when considered appropriate, or survey information will be documented on survey forms. Information that will typically be included on the survey maps or forms is:
- 1. Location of the measurement or sample.
- 2. Date and time of the .neasurement or sample.
- 3. The name of the surveyor, sampler or analyst.
- 4. Description and purpose of survey or sample,
- 5. Description of sampling equipment, including calibration dates.
- 6. Analysis date and time (if applicable).
- 7. The analytical error (if applicable).
- 8. Units of measurement or analysis.
- 9. Unique conditions pertaining to the survey or analysis.
All original survey data shall be retained and placed in PSC archives at the termination of the project. 4.5.2 RDAlolonical Survey Rengd At the completion of the decommissioning effort, a radiological survey report will be developed to document the findings and conclusions of the final survey. This report will provide the basis for securing approval for the termination of the 10 CFR 50 license. The radiological survey report will contain an overview of the radiological condition of the site and structures, a detailed presentation of the data in the form of tables and figures, and interpretation of results relative to the decommissioning release criteria, it will also describe the residual radioactivity in the remaining structures and systems to characterize the final facility and site radiological condition. 4.6 QUAllTY ASSVRANCE The objective of quality assurance, as applied to the final radiation survey plan, will be to ensure confidence in the sampling, analysis, interpretation and use of the data generated from the final survey. Quality assurance for the final radiation survey plan will be an integral part of the overall decommissioning QA plan and will be governed by Section 7 of this plan. 4.7 INDEPENDENT VERIFICATION A third-party independent verification of the final survey will be performed as an audit of the final survey plan. This independent 4-8
I l 6/28/91 PROPOSED DECOMMISSIONING PLAN l SECTION 4 verification will include selected measurements, sampling and analysis as required to confirm validity of the final survey. I The independent verification program will also require formal program development to remove possible judgmental factors and prevent skewing of the final results. The independent verification program will be of similar structure (although on a smaller scale) as the final survey plan. i a 4-9
~ __ _ A _ _ . _ . . _ - . _ . _ _ _ . _ _ . _ . . _ - _ _ .
_ _ . ~ _ . _ _ . - . - _ - - _ _ - _ .._
s ., - - PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 4
4.8 REFERENCES
FOR SECTION 4
- 1. NRC Regulatory Guide 1.86: "Termin8 ion of Operating License for Nuclear Reactors." June 1 '4.
- 2. NRC Memorandum P.B. Erickson (NRC) to Seymour H. Weiss (NRC);
Subject:
" Summary of Meeting with Public Service Company of Colorado (PSC) To biscuss Preliminary Decommissioning Plan of June 30, 1989," dated August 24, 1989.
- 3. NRC letter, John c. S' z (NRC) to Dr. Roland A. Finston i, Stanford), dated March 4 1981.
NRC letter, James R. Miller (NRC) to Dr. Roland A. finston 4. (Stanford), dated April 21, 1992. b
- 5. NUREG-0586: " Final Ceneric Environmental Impact Statement on Decommissioning of Nuclear Facilities." August, 1988.
.6. "Posidual Radioactive Contamination from NUREG/CR-5512:
Decommissioning." ' aft Report. January, 1990
- 7. NRC Circular 81-07: " Control of Radioactively Contaminated Material ." May,1981.
- 8. NRC Information Notice 85-92: " Surveys of Wastes Before Disposal from Nuclear Reactor Facilities." December,1985.
- 9. NUREG/CR-2082: " Monitoring fo- Compliance with Decommissioning Termination Survey Criteria." (ORNL/HASRD-95). June, 1991.
4-10
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 5 SECTION 5 DECOMMISSIONING FIXED PRICE CONTRACT AND FUNDING PLAN 5.1 DECOMMISSIONING CONT 0ACT As noted in Section 2.1, PSC has selected the DECON option for early decontamination, disaantlement, and decommissioning of the radioactive portions of the Fort St. Vrain Nuclear Generating Station. In order to accomplish this project, PSC released a Request for Proposal to several highly qualified companies for the purpose of receiving competitive bids on the project. Four qualified bids were received and, based on a thorough evaluation for technical and financial acceptability, PSC selected a- project team of Westinghouse and MK Ferguson to decommi:.sion Fort St. Vrain, with Westinghouse as the lead contractor. PSC and the Westinghouse team have reached agreement on to a final contract to perform the decommissioning work. The selection of the Westinghouse team as a result of the the competitive bid process resulted in a tot al contract price of
$137,129,000 for the decommissioning of Fort St. Vrain, inclusive of escalation and PSC expected costs. A detailed cost e.;tiraate was prepared which provides a detailed breakdown of these cost . This detailed cost estimate was submitted to the NRC in Reference l Figure 1.3-1 of Reference 1_ provides a summary breakdown of' the project costs based on the major decommissioning activities.
The use of a firm fixed price contract greatly reduces the level of uncertainty in the decommissioning cost. By use of the competitive bid process, an accurate method has been utilized to determine the real cost for decommissioning, based on the identified scope of work and assumptions. The bid process and resulting contract commits both PSC and the Westinghouse team for the project scope and cost. Certain restrictions and limitations exist when only a cost estimate has been prepared as a basis for evaluating decommissioning costs and as a basis for the decommissioning funding pl an. A cost estimate is limited in tSat it is only a study to determine reasonable estimates of individual costs and involves no commitment on the part of the cost estimator to meet the estimate during the actual performance of the work. A firm fixed price contract goes beyond this phase, in that a contractor is bound under a contractual obligation to perform this established scope of work at the price they have bid. 5-1 l
PROPOSED DECOMMISS10NING-PLAN 6/28/91 SEf110N 5 Receiving bids from four qualified bidders was equivalent to iaceiving four independent cost estimates. Since each bid utilized a different decommissioning methodology, this approach exceeds any regulatory guidance for financial assurance and is beyond that re4uired by the Decommissioning Rule. In evaluating the four bids, detailed assessments of the actual decommissioning work and methodology were conducted to ensure that the bidders had adequately identificd and accounted for the work to be performed. Detailed evaluations and cross comparisons were also conducted to ensure that the bidders had adequately addressed technical. support requirement, project management and control, radiological _ waste handling, radiation protection, facilities and support requirements, quality assurance and project documentation and closcout. Areas of uncertainty were identified and clarified with the bidders, including evaluations of pricing contingencies
- regarding waste volumes, contamination levels, etc.. The use of i- this competitive bid process, the high quality of the responses
( received, and the detailed bid evaluations that were conducted, provides significant confidence in the cost estimate as well as the overall decommissioning approach and the work scope. Therefore, PSC is confident that all major tasks have been identified and included r within the Westinghouse team fixed price contract. I l 5.2 MAJOR ASSUMPTIONS, BASES, AND SCOPE OF FIXED PRICE CONTRACT The following information is provided to identify the basis of the
<ixed price contract between PSC and the Westinghouse team to decommission Fort St. Vrain. A detailed breakdown of the Westinghouse team proposed scope of work is provided in Appendix I of this plan. The following major work activities and necessary l
support activities will be performed: i
- _ l. Decontaminate in place, and/or remove and decontaminate, and/or remove and dispose of the contaminated and activated l materials -.inside the PCRV and those that form the PCRV L structure.
L l 2. Decontaminate in place, and/or remove and decontaminate, and/or remove and dispose of the contaminated portions of the plant systems outside of the PC,iV.
- 3. Survey and cleanup the site as required, including t' e evaporation ponds and effluent blowdown flow paths.
5-2
i l l 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 5 Decontamination and decommissioning activities will be performed to the extent necessary to decontaminate all radioactive portions of the plant to the final release criteria specified in Section 4.2 of this plan. All other materials remaining as part of the PCRV structui.s, in the systems outside of the PCRV and on the site after the final radiation survey will be confirmed to be below these release limits and will remain on-site. As noted in Section 2.4, PSC is responsible for overall project management and licensing interface with the NRC. Major PSC responsibilities (and associated costs) include:
- 1. Overall control of the project
- 2. Access control
- 3. Radiation protection overview
- 4. Quality assurance overview
- 5. Licensing coordination
- 6. Operation and maintenance of required plant :ystems
- 7. Responsibility for the final independent radiation survey
- 8. Engineering configuration control overview The following are major assamptions included in the basis of the firm fixed prico and tne detailed cost estimate:
- 1. The current facility design and layout is as described in Section 2.2 and shown in Figure 2.2-1, and no major modifications are anticipated.
- 2. Radionuclide inventories, activation analyses, and estimated dose rates are as described in Section 3 of this plan.
- 3. PSC will supply utilities to the contractor, including electric power and water, and the cost for these utilities is included in the cost estimate.
- 4. No mixed waste or contat.inated asbestos exists.
- 5. Burial charges are based on the current disposal rates in affect at ' the ' Beatty NV disposal site until the end of 1992. A contingency has been added for burial of radioactive waste at the Richland -WA disposal site after l 1992 L 6. No cost allowances were included for major schedule delays
! caused by uncontrollable and. unforeseen events. Appropriate contingencies are inci t.ded to account for project uncertainties. l 7. Existing plant equipment will be utilized when determined l to be cost effective and technically sound to operate and L maintain. 4 l 53 I
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.l PROPOSED DECOMNISS10NING PLAN 6/28/91 SECTION 5 L
major project scope changes will be made according to the changes experienced.- Based on these annual reviews of decott.mi ssioning price, the decommissioning funding plan will also be reviewed and revised accordingly. Since the project is scheduled for completion within 39 months after commencement 'of physical dismantlement and decommissioning activities, adjustments will be made as frequently as deemed necessary for successful funding of the project. The NRC will be informed of .any changes exceeding (plus or minus) 15 percent to either the decommissioning price or the decommissioning funding plan. i. l 5-6
6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 5
5.7 REFERENCES
FOR SECTION 5
- 1. PSC letter, Crawford to Weiss, dated June 6, 1991 (P-91198)1 lu_bject: " Fort St. Vrain Decommissioning Cost Estimate".
- 2. USNRC Draft Regulaton! Guide DG-1003, " Assuring the Availability of funds for Decommissionirig Nuclear Reactors," September 1989.
- 3. PSC letter, Crawford to Weiss, dated February 15, 1990 (P-90039), i I
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PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 5 l I I INTENTIONALLY LEFT BLANK i l i. l i 1 5-8
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6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 7 SECTION 7 DECOMMISSIONING QUALITY ASSURANCE PLAN 7.1 POLICY STATEMENT Public Service Company of Colorado (PSC) will establish and implement a Quality Assurance Plan for the Fort St. Vrain (F3V) Decommissioning Project. This Quality Assurance Plan is based on the requirements of 10 CfR 50 Appendix B as they apply to decommissioning activities and is responsive to other applicable regulatory requirements, and industry codes and standards. The goals of the Quality Assurance Plan are to provide protection of the health and safety of the project personnel and the public, and to comply with regulations and commitments made to the NRC, including the control of personnel exposure to radiation, control of radioactive material, control of radioactive material shipment, and final radiological survey. Project procedures shall provide for compliance with appropriate regul atory, statutory, and license requirements. - Specific quality assurance requirements and organizational responsibilities for implementation of these requirements shall be specified. compliance with this plan and project procedures is mandatory for personnel with respect to Fort St. Vrain decommissioning activities which may affect quality and the health and safety of project personnel and the general public. Personnel shall, therefore, be familiar with the requirements and responsibilities of the plan that are applicable to their individual activities and interfaces.
7.2 INTRODUCTION
This Quality Assurance Plan is applicable to and is structured to assure that the regul atory requirements as ider'ified in the Proposed Decommissioning Pl an , the requirements of the Decommissicning Technical Specifications (DTS), the requirements of the Radiation Protection Program, the packaging and shipping of raoloactive materials, and the final radiation survey are conducted in a controlled manner. l l l _ 7-1
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 7 7.3 ORGANIZATION 7.3.1 Einen1 The Quality Assurance organizations of PSC and the Westinghouse team have the authority and organizational freedom to identify quality problems; to take action to stop unsatisfactory work and control further_ processing, delivery, installation or use of nonconforming items; to initiate, recommend, or provide solutions; and to verify implementation of solutions. The persons and organizations performing quality assurance functions report to a management level that assures the required authority and organizational freedom are provided, including sufficient independence from cost and schedule. The individuals assigned the responsibility for assuring effective execution of any portion of the Quality Assurance Plan have direct access to the levels of management necessary to perform quality assurance functions. 7.3.2 PSC Oraanization i The PSC Organization is explained in Section 2.4 of this decommissioning plan. Section 2.4 provides organizational charts, together with a summary of the authority and duties of key decommissioning staff members. PSC has overall responsibility for the Quality Assurance (QA) Plan implementation, and is responsible y for verifying the effective execution of the plan. PSC performs oversight of the Westinghouse team implementation of the plan through reviews, audits and monitoring activities (surveillances), l-7.3.3 'Westinahouse Team Oraanization l E 7.3.3.1 W ality Asjuranca ( The Westinghouse Nuclear and Advanced Technology Division 1HNATD) l Quality Assurance Manager reports directly to the WNATD General Manager and ultimately to the Energy Systems Business Unit Vice l Pre.ildent and Ger.eral Manager to ensure the independence of the QA function. The Quality Assurance Manager reports to the Westinghouse
, Project Director' for_ administrative direction and implementation of the Quality Assurance Plan. The Quality Assurance Manager and the Project Director are responsible for assuring effective execution of the Quality Assurance Plan. MK-ferguson personnel will work under the HNA10_ QA Plan during decommissioning activities and therefore the HNATD QA organization will apply. Westinghouse Scientific Ecciogy Group (WSEG) will implement their NRC approved 10CFR71, 7-2
i l 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 7 subpart H, QA Plan, which includes a completely independent QA organization, for packaging and transporting radioactive material. 7.3.3.2 Kev Decommissioning _T,,taff Members The Westinghouse decommissioning staff is explained in Section 2.5 of this decommissioning plan. Section 2.5 provides organizational charts, together with a summary of the authority and duties uf key merrbers . 7.4 QUALITY ASSURANCE PLAN 7.4.1 General lleauirements
- 1. The Quality Assurance Plan shall be:
- a. Documented by written procedures,
- b. Carried out throughout the decommissioning project in accordance with those procedures.
- 2. The plan shall include identification of the fallowing:
- a. The structures and activities to'be covered,
- b. The major organizations participating in the plan, together with the designated functions of these organizations.
- 3. The _ plan -shall provide control over ectivities affecting quality and the health' and safety of project personnel and the general public.
'4. Activities affecting quality shall -be accomplished under suitable controlled conditions. Controlled conditions i ncl ude -- the use of appropriate equipment, suitable-
, environmental conditions for accomplishing the activity, I and_ assurance that all prerequisites for - the given activity have been satisfied.
- 5. The plan shall take into account the need for special controls, processes, test equipment, tools, and skills to attain the required quality, and the need for verification of satisfactory implementation.
L 6. The plan shall provide for indoctrination and training of personnel performing activities affecting quality to assure that suitable proficiency is achieved and maintained. L 7-3 L
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 7
- 7. The adequacy and status of the plan shall be regularly reviewed.
- 8. Management of those organizations participating in the plan shall regularly review the status and adequacy of. that part of the plan which they are implementing.
- 9. The plan will be implemented for the final radiological survey to assure confidence in the sampling, analysis, interpretation, and use of data generated, it will apply to all aspects of the survey plan, from personnel-qualifications-to sampling and field measurements, handling and storing samples, sample reduction, reporting, and records turnover.
7.4.2 General Description ;
- 1. 'This Decommissioning Quality Assurance Plac has been l established to govern those activities that may affect the -l quality of the project, including the health and safety of the project personnel and the general public.
- 2. This Decommissioning Quality Assurance Plan shall utilize !
the following documents to meet its objectives. l
- a. The Westinghouse Fort = St. Vrain Decommissioning Project Quality Assurance Plan which provides the '
details of the Quality Assurance Plans and procedures that will be utilized by each Westinghouse organizational team member,
- b. Westinghouse required procedures at the project implementing level,
- c. PSC oversight process and procedures 7.4.3 Kestinohouse FSV Decommissionino Proiect Ogality Plan
- 1. The plan shall define the QA plans and procedures that will be used by each Westinghouse organizational team member.
- 2. The plan shall be issued and approved by -Westinghouse and PSC.
- 3. All changes to the Project Quality Plan shall be governed by measures commensurate with those applied to the original issue.
7-4 l- _ -
6/28/91 DECOMMISSIONING PLAN SECTION 9 SFCTION 9 DECOMMISSIONING EMERGENCY RESPONSE PLAN
9.1 INTRODUCTION
AND REGULATORY BASIS The fort St. Vrain reactor was rc.manently shut down in August 1989. PSC evaluated the credible defueling accidents and determined that the required level of emergency response capability was significantly less than that for the operating reactor. As a result, PSC prepared a Defueling Emergency Response Plan, with a significantly reduced level of emergency response, and submitted it to the NRC in June 1990. The credible accidents for dismantlement and decommissioning operations have been evaluated in Section 3.4 of this plan. This evaluation indicates that reduced emergency response capabilities are appropriate for the Decommissioning Emergency Response Plan. As the dismantlement and decommissioning proceeds, various tasks and milestones will be completed such that the asscciated postulated accident scenario will no longer be considered credible. With the reduction in the accident scenarios and resultant consequences, the Decommissioning Emergency Response Plan may be revised to eliminate selected emergency response capabilities. 9.2 DECOMMISSIONING EMERGENCY RESPONSE PLAN SCOPE 9.2.1 Maximum Emeraency Action level (EAll - AL GI Based on an evaluation of the credible dismantlement and decommissioning accident;, the accident consequences have been compared with the emergency classification tables contained in FSV procedures and the guidance contained in NUREG-0654 Appendix 1
-(Emergency Action Level Guidelines for Nuclear Power Plants) and the EPA Protective Action Guidelines.
The worst case dismantlement and decommissioning accident has been identified to be the postulated tractor / trailer fire from the rupture of the diesel fuel tanks in which the trailer is conservatively loaded with 230 side spacer ' blocks (boron pins removed). These side spacer blocks have the highest potential releasable quantity of radioactivity. Section 3.4.6 provides a
- l. detailed analysis of _ this accident scenario.
Based on the-conservative assumptions for the percentage of graphite l oxidation, radionuclide inventory and release, the radiological consequences to an individual standing at a point on the EPZ, which 9-1
~_ . __ ___ . - _ _ _- - _
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 9 is a minimum distance of 100 meters from the Rcactor Building, the fuel Storage Building and the Waste Compactor Building, is a whole bcdy and lung dose of 121 mrem and 215 mrem, respectively. Therefore, no accident condition will exceed the NUREG-0654 guidelines for an ALERT. 9.2.2 EPZ RestricteJ1 to Within 100 Meters of the Reactor Buildina Gased on evaluation of the dismantlement and decommissioning accident consequences in Section 3.4 of this plaa, there will be no credible means for offsite exposures to exceed a small fraction of the EPA Protective Action Guidelines or to create the need for either a plume exposure or an ingestion pathway planning _ zone. Therefore, the EPZ boundary is designated as the following area: To the east, north and northwest, the EPZ boundary follows 1 the Protected Area Fence. ! To the south, the EPZ boundary follows the southern i boundary of the switchyard and the Fire Training Facility I To the southwest, the EPZ boundary follows the plant side of county road 19-1/2. The EPZ boundary is sufficient to control access to the onsite EPZ and will allow unrestricted use of Weld County Road 19-1/2. Perimeter monitoring of this onsite EPZ will be performed following any potential -radiological release as required by the emergency situation to confirm that there will be no offsite radiological releases above those determined in the analyses. 9.2.3 Offsite Emeroency Response by State / Local Governments Eliminated Section 1.4 of_ NUREG-0696, " Functional Criteria for Emergency Response Facilities", provides- the following guidance regarding activation of ERF's:
- 1. " Activation of the onsite TSC and OSC is optional for a Notificatton of Unusual Event emergency class, and is required for ALERT and higher classes.
- 2. Activation of the nearsite EOF is optfonal for Notification of Unusual Event and Alert emergency classes, and is required for Site Area Emergency and General Emergency classes."
9-2
.- - ~ _. . . -
PROPOSED DECOMMISSIONING PLAN 6/28/91 SEC110N 9 Additionally, PSC will continue to perform initial notification of the State and local governments (via the Weld County Communications Center and the Colorado Department of llealth), as well as provide l both organizations with periodic status updates. 9.2.5 Dissemination of Emeroency Phn Information to local i Residents The need for annual dissemination of basic emergency planning l information to the public located within the five-mile radius (the approved EPZ for operations) will no longer be necessary and will not be a requirement of the Decommissioning Emergency Response Plan. 9.2.6 Need for the Offsite Emeroency Warn 4 1a System The Emergency Warning System, which consists of tone-alert radios and the Platteville siren to warn local residents of protective l actions, will no longer be maintained as part of the Decommissioning
- Emergency Response Plan.
9.2.7 The PCC and TSC Will be Consolidated Under the Control of the TSC Director (Emeraency Coordinator) The onsite emergency response manpower requirements have been evaluated as a result of the reduced consequences of remaining accidents and the elimination of offsite ERF's, and it has been determined that onsite manpower may be reduced and onsite ERF's consolidateo. Changes will be made to ERF staffing assignments to reflect the reduced communication requirements with offsite ERF's. Based on this evaluation, the TSC and Personnel Control Center (PCC) functions will be consolidated in the TSC under the direction of the TSC Director. Consolidating the PCC into the TSC is acceptable for l the following _ reasons: (1) onsite radiological consequences are l-significantly reduced, which should allow controlled access to all areas in and around the Reactor Building within the 100-meter EPZ; (2) elimination of offsite communication responsibilities reduces minimum staffing requirements; (3) based on minimum staffing levels , for an ALERT, this consolidation will not cause congestion in the l TSC; and (4) consolidation will _ allow the Emergency Coordinator (TSC l Director) direct control of onsite activities, including field and l in-plant surveys, search and rescue, and plant recovery teams dispatched to control-and mitigate accident consequences. 1 As the onsite Emergency Coordinator, the TSC Director has always been ultimately responsible for the function of the PCC, a'though 9-4 l l
6/28/91 DECOMMISS10NING PLAN SECTION 9 the PCC Director performed oversight and implementation of these duties. In consolidating the PCC into the TSC, the need for a PCC Director will be eliminated, communications will be handled more directly and of fectively, and the potential for miscommunication will be decreased. With the decrease in severity of remaining credible radiological emergencies, the demands on the TSC Director will be proportionally This in turn will allow the TSC Director the opportunity decreased. to assume direct oversight of duties previuusly performed by the PCC without decreasing the effectiveness of the emergency response organization. Consolidation of the teams into the TSC will allow the TSC Director more direct control of corrective actions, while the reduction _ in minimum staffing will prevent this consolidation from interfering with the onsite command and control responsibilities of the TSC. 9.2.8 The EPZ Field Team Based -on the revised accident and radiological consequences, there will be negligible offsite radiological consequences. The EPZ field monitoring team will continue to conduct radiological surveys within the decommissioning EPZ (see Section 9.2.2). There is no need for State radiological monitoring. l 9.2.9 Distribution of KI Tablets Both Onsite and Offsite With the EPZ limited to onsite areas and with negligible short term offsite radiological consequences, there is no longer any need to stock KI tablets. Additionally, all I-131 in the primary coolant nas undergone well over fifty half-lives of radioactive decay since-l the permanent shutdown of Fort St. Vrain and would be non-existent in any future radiological release from Fort St. Vrain. 9.-2.10 Notification of Conversion Work Force Workers involved with conversion efforts will receive information on
- j. how to respond .in the event of . activation of the Decommissioning Emergency Response Plan. Response actions will be initiated to l
ensure that exposures of conversion workers are kept ALARA. 9.2.11 f_qrmal Recoverv Oraanization and Procedures Revised The reactor will have been shut down and all fuel removed. The formal recovery procedures required by regulations will be revised to reflect: (1) reduction in the number of cperating systems needed 9-5 l
.- . - .. - - -. . . . . . . .- .. .- ~- .
'l PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 9 to maintain any- special conditions for decommissioning;- (2) reduction in the severity of possible accident consequences and possible radiological emergencies; (3) reduced minimum staffing of the onsite emergency response organization.
(I l-9-6 i j
4 6/28/91 PROPOSED DECOMMISSIONING PLAN SECTION 10 l l SECTION 10 DECOMMISSIONING FIRE PROTECTION PLAN
10.1 INTRODUCTION
l The Fort St. Vrain Decommissioning Fire Protection Plan (D/FPP) will I be structured to protect the safety of the decommissioning workers, minimize the risk and consequences of fire ' damage to the property and minimize the release of radioactive contamination to the public l due to a fire involving radioactive materials in order to maintain l off-site doses within the guidelines of 10CFR100. Radiolegical ! consequences of fires involving the release of radioactive material are discussed in Section 3.4.6. l The p/FPP will include the following as required to meet the objectives stated above: fire prevention, fire detection and fire suppression. The D/FPP will be under the jurisdiction of the PSC Decommissioning Engineering Manager and will be administered within the Westinghouse team organization, as appropriate. The D/FPP will identify the appropriate portions of the existing fire detection, fire barriers, ~ tire deors and automatic water suppression systems that will be maintained during decommissioning. The existing fire pumps, fire water distribution system and associated hose stations will be maintained.
- Portable fire extinguishers will be selected, located and maintained in the general work and storage areas. An inventory of fire extinguishers will be maintained to be issued to those on fire watch duty. permanent fire protection features will be described in the D/FPP. Work areas will be inspected to assure that housekeeping is being attended-to, combustibles are properly stored and unnecessary combustibles are not being allowed to accumulate in work areas.
The D/FPP will required the implementation of a Hot Work Permit system for all decommissioning activities involving an ignition
- source such as torch cutting or spark prooocing tools. The Hot Work l Permit system will include pre-work inspection to assure the area is '
L clear of any unnecessary combustibles, the assignment of a trained fire watch with appropriate extinguishers and a pcst-work watch period concluding with inspection of the area. The Hot Work system will address the elements of fire prevention, fire detection and l fire eJppression in conformance With the requirements of the D/FPP. The personnel assigned to fire watch duty will be trained on the
- 1. duties and responsibilities of performing fire watch functions including the use of firc extinguishers. The fire watch will axtinguish any fires judged to be within the capability of the 10-1 :
l
PROPOSED DECOMMISSIONING PLAN 6/28/91 SECTION 10 available equipment and will report any fire to the PSC Shift Supervisor. The Shift Supervisor or designee will direct the fighting of any fire where assistance is req tired and will obtain the services.of off-site fire departments as the situation dictates. Y I 2 I i-
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