ML20085G718
| ML20085G718 | |
| Person / Time | |
|---|---|
| Site: | 05000047 |
| Issue date: | 10/31/1991 |
| From: | EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | |
| Shared Package | |
| ML20085G716 | List: |
| References | |
| EGG-WM-9184, EGG-WM-9184-R01, EGG-WM-9184-R1, NUDOCS 9110280004 | |
| Download: ML20085G718 (62) | |
Text
{{#Wiki_filter:g g Rovlslon 1 yg gg October 1991 US Army Corps of Engineers Toxic and Hazardous e Materials Agency DECOMh1ISSIONING PLAN FOR U.S. ARMY h1ATERIALS TECHNOLOGY LABORATORY RESEARCH REACTOR h[.k ~ ' i ; c '. i i4? r- t or< 1Ht,MA ;:orm 45.1 Jul 90
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l NOTICE The views, opinions, and/or findings contained in this report are those cf the author (s) sad should not be construed as an official Department of the Army position. policy, or decision, unless so itsignated by other *. documentation. 3 The use of tradensmes in this report does not constitute an official endorsement or approval of the use of such commercial products. This report may not te cited for purposes of advertisement. 9 11 i L J
EGG W!O9184 l Revision 1 Octot.er ;991
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DECOMMISSIONING PLAN FOR U.S. ARMY MATERIALS TECHNOLOGY LABORATORY RESEARCH REACTOR Prepared by EG&G Idaho, Inc. Idaho liational Engineering l.aboratory As Part of Work for Others Project tio, 88845 Prepared for the United States Army Toxic and Hazardous Materials Agency
. Base Closure Division Aberdeen Proving Ground. Maryland 21010 and for the U.S. Department of Energy Idaho Operations Office Under DOE Contract No. DE.AC07-761001E70 s
r l i EXECUTIVE
SUMMARY
l This DemniniortnalintIDELf0LUAM1_AtIn1MahtialLinhn1991
, Lahernery_hElg at Watertown, Massachusetts, specifies that the reactor !
facility will be decommissioned by decontamination. The Army decided upon total dismantlement as the pref erred alternative f or reasor,$ explained in this document. However, this Op describes decommissioning of the reactor f acility by continuous dismantlement as required in order to decontaminate by removing contaminated materials. This will accomplish decommissioning and lead to the release of the facility for unrestricted use. 1he facility to be deccmmissioned includes the containment building (fluilding 100), piping in the reactor facility laboratory (lluilding 97), an 4 underground pool water retention tank and remnant components of the secondary [ coolant system. The following tasks will be undertaken in order to complete the decommissioning of the reactor. Some of these may be performed in parallel. ! first, the Decommissioning Contractor will remove any auxiliary structures, which will include Cir. tern 242, the piping in Building 97, and the secondary l coolant-system. Secondly, the Contractor will remove the components and systems from the reactor building itself. This will involve removing the reactor pool inter' is, reactor pool liner, platforms, basement piping, basement sumps, the ganna facility and storage tubes, and the reactor pool. [ following completion of decommissioning, the termination survay will be-performed by the U.S. Army Environm9ntal Hygiene Agency. The survey will-N include the reactor l$uilding and the soil areas where excavations were periormed. A Quality Assurance plan will be developed and followed throughout the decommissioning process to ensure that the work is performed in compliance v U . . - . . - . . . - . . . - - . . - . ~ _ - . - _ . - . . - .
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p with the Decommissioning Plan, the Statement of Work, procedures, and other applicable specifications and requirements, in addition, the decommissioning program will be conducted in such a way as to be certain that the following regulations, guides, and standards are met: all applicable regulations from the Code of Massachusetts Regulations; pertinent portions of the U.S. Code of federal Regulations; applicable Army regulations; regulatory guides from the regulating agencies (e.g., the Nuclear ' Regulatory Commission and the Environmental Protectior uncy); and the standards set forth by certain institutions or technical societies (e.g., American Society for Testing and Materials, the International Commission on Radiological Protect lon, and the American National Standards Institute). The Occupational and Radiation Protection Programs for the AMTL Reactor decommissioning will consist of a set of policies, procedures, and instructions to protect workers, the general public, and the environment. The i Occupational and Radiation Protection Programs will provide occupational health, health physics, industrial hygiene, and safety elements. The Radiation Protection Program will include requirements to monitor radiation and radioactive materials, to control distribution and releases of radioactive materials, and to keep radiation exposure within 10 CFR 20 limits and at as-low as reasonably-achievable limits. The related Industrial Safety and Hygiene Program will be concerned primarily with protection against nonradioactive exposures and hazards and will be administered in accordance with regulations from the Occupatienal Safety and Health Administration. The decommissioning operations will be performed and managed by a decommissioning contractor. As the licensee, the AMIL Commander is . responsible for the overall decommissioning project and has authority in all associated matters, including-safety. The overall project management,- , however, will be acenmplished by the U.S. Army Corps of Engineers, New England Division. The AMTL will provide a quality assurance expert to ensure that the contractor is performins work in accordance with terms of the contract. vi , J
The estimated cost of the AMIL Reactor deconrnissioning project ranges between $4.3 am: $5.1 million. Both estimates are based on decontamination The lower estimate is based on an assumed void volume of 10% ;r, tha nachog. radioactive waste, and the higher cost estimate is based on an assumed void volume of 60%. The estimates also assume that the radioactive waste wculd be
** disposed of Letween 1 January 1992 and 1 January 1993, which would involve paying a state penalty surcharge of $120 per cubic foot of radioactive wt.ste.
The funding for this project will come from appropriate U.S. Army source; and will be distributed by the U.S. Army Corps of Engineers Military Programs (CEMP). i Y
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CONTENTS NOTICE ................................ ii EXECUTIVE
SUMMARY
, . . . . . . . . . . . . . . . . , , . ...... v ACRONYMS ............................... xv ,
CHAPTER 1, t3ACKGROUND AND MANAGEMENT PLAN . . . . . . . . . . . . . . . . 1-1
- 1. INTRODUCTION .................. ......... 1-1 1.1 Summary Description ...................... 1-1 1.1.1 Reactor facility Description .............. 1-1 1.1.2 Reactor Usage During Licensed Period .......... 1-1 1.1.3 Comparison of Considered Decommissioning Alternatives . 1 14 1.1. 4 Major Advantages of the Preferred Alten ative . . . . . 1-16 1.1.5 H;jor Tasks and Schedules . . . . . . . . . . . . . . . 1-16 1.1.6 Quality Assurance Plan . . . . . . . . . . . . . . . . 1-17 1.1.7 Contractor Participation ... .. . . . . . . . . 1-18 1.1.8 Termination Radiation Servey Plan . . . . . . . . . . . 1-19 1.2 facility Operating and Post 'Jperating History . . . . . . . . 1 19 1.2.1 Reactor Power History and Experiments . . . . . . . . . 1-19 1.2.2 Leaks in Reactor Systems ......... ..... . 1-21 1.2.3 Unscheduled Shutdowns . . . . . . . . . . . . . . . . . 1 22 1.2.4 Unplanned Release from Cistern 242 .......... 1-22 1.2.5 Reactor Operation Summary . . . . . . . . . . . . . . . . 1-23 1.2.6 Post-0perating History ............... . 1-23 1.3 Current Radiological Status of f acility . . . . . . . . . . . 1-25 1.3.1 Radiological Characterizat ion . . . . . . . . . . . . . 1 25 1.3,2 Neutron Activation Analysis ,,,... . . . . . . 1-47 1.4 Estimated Personnel Dose . . . . . . . . . . . . . . . . . . . 1 52 1.5 Estimated Volume of Radioactive Material to be Removed . , . 1-52 1.6 Decommissioning Alternative . . . . . . . . . . . . . . . . . 1-55 ,
a 1.7 Decommissioning Organization and Responsibilities . . . . . . 1 55 1.8 Regulations, Regulatory Guides, and Standards .. . . . . . 1-58 l.8.1 Applicable Regulations ..... . . . . . . . . . . 1-58 1.8.2 Regul atory Guides . . . . . . . . . . . . . . . . . . 1 -60 1.8.3 Standards . . . . . . ..... . . . . . . . . . . . 1-61 viii
l.8.4 Applicable Army Requirements and Regulations 1-61 1.8.5 Informal Guidance and lechnical Rcports . . . . . 1-62 1.8.6 Permits / Licenses Covering the AMIL Reactor . . . . 1 62 1.9 Training and Qualifications . . . . . . . . . 1-63 1.9.1 Training Program Descriptions . . . . . . 1-63 .. 1.9.2 Administration and Recordkeeping . . . . 1-65 CHAPTER 2, OCCUPA110NAL AND RADI A110N FR01[CT10N PROGRAMS . . . . . 2-1
- 2. INIRODUCTION . . . . . . . . . . . . . . . . . . . . 2-1 2.1 Radiation Protection Program . . . 2-2 2.1.1 Personnel . . . . . . . . . . . . . . . . . . . . . . . 2-?
2.1.2 Notices, Instructions, and Reports to Workers, inspections . . . . . . . . . . . . . . . 2-3 2.1.3 Training . . . . . . . . . . . . . . . . . . . . . . . 2-3 2.1.4 Administrative and Raulological Controls . . . . . . 2-5 2.1.5 Radiation Protection facilities, Instrumer.tation, and Personal Protective Equipment . . . . . 2-14 2.2 Industrial Safety and Hygiene Program . . . . . . . . 2-22 2.2.1 Personnel . . . . . . . . . . . . . . . 2-22 2.2.2 Training .. . . . . . . . . . . . . . . . . . . ?-23 2.2.3 Administrative and Work Practice Controls . . . . 2-24 2.2.4 Operational Activities . . . . . . . . . . . . . 2-29 2.2.5 Personal Protective Measures . . . . . . . . . 2-31 2.2.6 Excavations . . .. . . . . . . . . . . . . . . . 2-31 2.3 Contractor Assistance , , . . . . . . . . . . . . . . . 2-33 CHAPTER 3, DECOMMISSIONING TASKS, SCHEDULE, C051, AND FUNDING . . . 3-1
- 3. INTRODUC110N . . . ... . . . . . . . . . . . . . . . . . . . 3-1 3.1 Site Preparation . . .. . ... . . . . . . . . . . . . . . 3-4 3.1.1 Building 100 . . . . . . . . . . . . . . . . . . . . 3-4 3.1.2 Building 97 . . . . . . . . . . . . 3-5 3.1.3 Auxiliary Structures ... . . . . . . . . 3-5 3.1.4 Miscellanecus . . . . . . . . . . . . . 3-5 3.1.5 Setup Trailers and Staging Areas . . . . . 3-6 '
3.1.6 Establish Barriers . . . . . . . . . . 3-6 3.2 Project Manacement . . . . ... . . . . . . . . . 3-7 3.2.1 Staten._nt of Work . . . . . . 3-7 3.2.2 Decommissioning Contract . . . . . . . . . . 3-7 3.2.3 Onsite Monitoring . . . . . . . . . . 3-8 ix
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1.2.4 Progress Reports ... . . . . . . . . . . . . . 3-8 3.2.5 linal Report .. ........ . . . . . . . . 3-9 3.2.6 Artifacts Stored / Displayed . . . . . . . . . . . 3-9 3.3 Auxiliary Structures Removal . . . . . . . . . . . . 3-9 3.3.1 Cistern 242 . . . . .... . . . . . . . . . . 3-9 3.3.2 Bi '1 ding 97 Piping .. . . . . . . . . . . . . . 3-12 . 3.3.3 5 idary Coolant System . . . . . . . . . . . 3-12 3.4 Reactor Building Decontaminat ion . . . . . . . . . . . . . . 3-14 ,, 3.4.1 Pool Internals and Annulus . . . . . . . . . . , 3-14 3.4.2 Pool Liner . .. .. . . . . . . . . . . 3-15 3.4.3 Platforms . .. .. . . . . . . . . . . . . . . . 3-15 3.4.4 Basement Electrical . . . . . . . . . . . . . 3-18 3.4.5 Basement Piping and Louipment . . . . . . . . . 3-20 3.4.6 Basement Sumps . ..... . . . . . . . . . . . . . . 3-?O 3.4.7 Gamma facility and Storage lubes . . . . . . 3-23 3.4.8 Pool . . . . .. .... . . . . . . . . . . . . . . 3-23 3.4.9 Building Internal Decontamination . . . . . . . . . . 3-24 3.5 Termination Survey . . . . . . . . . . . . . . . . . . . . . 3-24 3.5.1 Phase I . . . ..... . . . . . . . . 3-25 3.5.2 Phase 11 . . .... . . . . . . . . . . . 3-25 3.6 Backfill and Grade . .. . . . . . . . . 3-25 0 3.7 Estimated Schedule and Cost ... . . . . . . . . . . . . . 3 26 3.8 funding . . . . . . .. . . . . . . , . , 3-26 CHAPTER 4 SECURITY . . . . . . . . . . . . . . . . . . 4-1
- 4. INTRODUCll0N . . . . . . . .... . . . . . . . . . . . . . 4-1 4.1 Physical Security . ..... . . . . . . . . . . 4-1 CHAPTER 5 RADIOLOGICAL ACCIDENT ANALYSIS . . . . . . . . . . . . . . . 5-1 CliAPTER 6 RADIDACTIVE MATERIALS AND WASIL MANAGEMENT . . . . . 61
- 6. INTRODUCTION . . . . . .. ...... . . . . . . . . . . , 6-1 6.1 fuel Disposal . . .... ..... . . . . . . . . . . 6-1 ,
6.2 Liquid Radioactivr: Waste . . . . . . . . . . . . . 6-1 6.3 Solid Radioactive Waste . . . . 6-2 a X N. _ _ _ _ _ _ . . _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
6.3.1 Packaging . . ... . . . . . . . . , 6-2 6.3.2 lemporary Storage of Radioactive Waste . . . . . . . 6-3 6.4 Ventilation System . . .... . . . . . . . 63 6.6 Waste Classification . ... . . . . . . . . . . . . . . 6-4 6.6 Shipping Radioactive Wastes ................ . 64 CHAPlER 7 TECHfilCAL Af4D ENVIRONMLt41AL SPEClfICA110t45 . . . . . . . . 7-1
, 7. IN1RODUC110N ..... ........ . . . . . . . . . . 7-1 7.1 Health and Safety limits .. . . . . . . . . . . . 7-1 7.1.1 External Exposure . ... . . . . . . . . . . . . 7-1 7.1.2 Internal Exposure . . . . . . . . . . . . . 7-2 7.1.3 Concentration of Airborne Radioactive Material in Restricted Areas ... , . . . . . . . . . . 7-?
7.1.4 Concentration of Airborne Radioactive Material in Unrestricted Areas ........ . . . . . 7-2 7.1.5 Concentration of lionradioactive Substances in Restricted Areas .... . .. . . . . . 7-2 7.1.6 Concentration of tionradioactive Substances in Unrestricted Areas .... . . . . . , . . . . . 72 7.1.7 lioise levels .. ........... . . . . . 7-2 7.1.8 As low as Reasonably Achievable (AlARA) . . . . . . . 7-3 7.2 Surveillance Requirements . . . . . . . . . . . . . . . 7-3 7.2.1 lhe Dosimeter Program . . . . . . . . . . . . . . . 7-3 7.2.2 The Routine Swipe Program . . . .. . . , , . . . . . 7-3 7.2.3 The Routine Instrument Survey Program . . . . . . . 7-4 7.2.4 The Air Sampling and Monitoring Program . . . . . . . . . 7-4 7.3 Administrative Controls ... . . . . . . . . . . . 74 7.3.1 Administrative Controls During Decontamination . . . . 7-4 7.3.2 Responsibility .............. . . . . 7-5 7.3.3 Organization .. . . . . . . . . . . . . . 75 . 7.3.4 Records and Reports . . . . . . . . . . . . . . . . . 7-5 7.4 Engineering Controls . . . . . . . . . . . . . . . . . . . . . 7-7 CHAPTEP 8 PROPOSED TERMINA110N RADIA110N SURVEY PLAN , . . . . . . . . 8-1
- 8. Il41R00VCTI0t1 ........... . . . . . . . . . . 8-1 8.1 Background Soil .. . .... . . . . . . . . . . . . 8-2 8.2 Phase 1 .................... . . . . 8-3 xi
8.2.1 Reactor Bcilding Interior Survey ..... ...... 8-3 8.2.2 Radiological Characterization of h enches and Pits . . 8-4 8.2.3 [ valuation of Phase 1 Results . . . . .. . . .. 56 8.3 Phase 11 . . . ... .. . . ... .... . .. 8-7 8.4 Instrumentation ...... . . . . ... . ... 8-3 8.5 Data Documentation . . . ... . . ............ 8- 8 R[f ER[NCES FOR EN11RL DOCUMENT ..... . ........ . . . 8-12 ., FIGURES 1-1. location of Army Materials Technolegy Laboratory in Watertown, Massachusetts. . . . ..... . . . . ..... . . l2 1-2. Army Materials Technology Laboratory general site map. .. ... 1-3 1-3. View of Building 100, looking west. Building 97 is in the background. .............. .... .. .. 1-5 1 4. Reactor containment shell cross-sectional view .,. . .. 1-6 1-5. Reactor containment-shell floor plans. . ... . ... . . 1-7 1-6. Army Materials Technology Laboratory Research Reactor model. .. 19 l-7. Reactor core support . .. .................. 1-11 1-A. View of Cistern 242. looking north. Building 100 is in background .............. . ... . 1-12 1-9. Smear survey locations for reactor baseiaent. ......... . 1-27 l-10. Smear survey locations for reactor operating floor. .... . 1-28 1-11. Smear survey locations for reactor first platform. .... . . . 1 29 i 12. Smear survey locations for reactor second platform. . .. . 1-30 1-13. Smear survey results and contact radiation readings for reactor annulus fuel element storage . . . . . . . . ... . . . . . . 1-31 1-14. Survey results and contact-radiation readines for reactor annulus components .. . . . .. , ,.... . . 1 32 1-15. Smear survey results and contact radiation readings for reactor annulus floor ... . .. .. . . . . 1-33 xit
1 1 16, Contact radiation readings for internal components of reactor ves sel (View No. 1) . . . . . . . . . . . . . . . . . . . . . . , 1 35 1-17. Contact-radiation readings f or internal components of reactor vessel (View No. 2) . . . . . . . . . . . . . . . . . . . . . 1-36 1-18. Contact radiation readings and smear survey results for internal l components of reactor vessel (View No.3) . . . . . . . . . . . . 1-37 l i 1-19. Smear survey results for internal components of reactor vessel (View No. 1) . . . . . . . . . . . . . . . . . . . . . . . . . . 1 38
- 20. Smear survey results for internal components of reactor vessel (View No. 2) . . . . . . . . . . . . . . . . . . . . . . . . . . 1-39 i 1-21. Building 100 and Cistern 242 area sediment and soil sample locations ... .. . . . .. . . . . . . . . . . . . . . . . . 1-43 1 22. AMTL Reactor decommissioning organization relative to safety . . 156 3-1. Work breakdown structure (WBS) . . . - ..........7 . . 3 3-2. Isometric view of Cistern 242 at the AMTL Reactor site. . . . . . 3-10 3 3. Photograph of the secondary coolant pumps and pad (The secondary sump is beneath_the pad). .-. . . . . . . . . . . . . . . . . . . 3 13 !
l 3 4. Cross section of the AMTL Reactor building. . . . . . . . . . . . 3-16 3-5. Second platform in the AMIL Reactor building. . . . . . . . . . . 3-17 l 3+6. _ first platform in the AMTL Reactor building. . . . . . . . . . . 1 3 19 , 1 5 3-7. Plan view of 'the AMTL R0 actor main floor. . . . . . . . . . . . . 3-21 3 8. Plan view of the AMTL Reactor basement floor. . . . . . . . . . . 3 22 3-9. Critical Path Method (CPM) . . . . . . . . . . . . . . . . . . . 3-27 8-1. Sketch of-AMTL Reactor site after completion of decommissioning . . , . . . . . . . . . . . . . . . . . . . . . . . 8-5 t I s TABLES [- l-1. Comparison of considered alternatives . . . . . . . . . . . . . . 1 1-2. Results of isotopic gamma scan for annulus smear sample . . . . . 1-3? xiii
1-3. Summary of the-results of the AMTL Building 100 smear sampling . ,1-40 1-4. Summary of the results of the AMTL Building 100 radiation survey .1-41 1-5. Analytical results of manmade true positive gamma-emitting nuclides for March 1990 sediment and soil sampics . . . . . . . . 1-44 1-6. Principal gamma emitting nuclides from neutron activation . . , . 1-50 1-7. Atoms present for activation . . . . . . . . . . . . . . . . . . 1-50 1-8. -Estimated Radiation Dose for Decommissioning of the AMll Reactor . 1-53 , 1 9. Estimated Radioactive Waste . . . . . . . . . . . . . . . . . . 1 54 2-1. Administrative guidelines for radiation whole-body doses during decommissioning . . . . . . . . . . . . . . . . . . . . . . . . . 2-6 2 2._ Regulatory limits for radiation doses during decommissioning for a calendar quarter (mrem) .,. ..... ........... ?6 2 3. Ty.nical-radiation survey and monitoring instrumentation and equipment to be-provided by-the Decommissioning Contractor . . . 2-17 2-4 Mitigation and monitoring of hazards during decommissioning . . 2-32 3 1. Cost estimate summary for decommissioning the AMIL Reactor. . . 3-28 8-1. Typical survey instruments and detection capabilities for measuring soil surfaces ..................... 8-9 s t k xiv l i--_ __ _ __ _ _ . _ . _ _ _
ACRONYMS ACGill American Conference of Governmental Industrial Hygienists AHERA Asbestos Hazardous Emergency Response Act ALARA a s - l ow- a s - re a sonabl y- ac t'i e v a bl e AMC Army Material Command AMCCOM U.S. Army Armament, Munitions, and Chemical Command AMMRC Army Materials and Mechanics Research Center AM1i. U.S. Army Haterials Technology L.aboratory Research Reactor ANS American National Standard ANSI American National Standards Institute ARCHS Army Reactor Committee for Health and Safety ASTM American Society for lesting and Materials Be0 beryllium oxide CCP Contamination Control Point Clli certified industrial hygienist CMR Code of Massachusetts Regulations COE Corps of Engineers COR Contracting Officer's Representative CPM critical path method CSCA Controlled Surface Contamination Area DOT Department of Transportation OP Decommissioning Plan dpm disintegrations per minute HEPA high-efficiency particulate absorpt ion HP Health Physicist ICRP International Commission on Radiological Protection
!!4EL Idaho National Engineering Laboratory INPO Institute of Nuclear
- Operations xv
_ _ _ _ _ _ _ _ _ _ _____ _ _____ _ __ ______ J
ISHP Industrial Safety and Hygiene Program K0 Contracting Officer LLD lower limit of detection LSA low specific activity , NCRP National Committee on Radiation Protection and Measurement NED New England Division
- NRC Nuclear Regulatory Commission NVLAP National Voluntary Accreditation Program OSHA Occupational Safety and Health Administration PPE . Personal protective equipment QAE Quality Assurance Evaluator QAE/HP Quality Assu ance Evaluator / Health Physicist RAM radioactive material RC&SO Radiological Control and Safety Officer RCA radiation control area
-RCC Radiation Control Committee RE/HPS Radiological Engincor/ Health Physics Supervisor RFP Req 0est for Proposal RHWPs Radiation / Hazardous Work Permits RPO Radiation Protection Officer RPP -Radiation Protection Program-t j .RSM -Radioactive Shipment Manifest '
S0W Statement of Work ,
.TLV Threshold Limit-Values l
VSAEHA U.S. Army Environmental Hygiene Agency
- l. USATHAMA United States Army Toxic and Hazardous Materials Agency l
WBS~ Work Breakdown Structure xvi
CHAPTER ) BACKGROUND AND MANAGEMENT PLAN
- 1. INTRODUCTION This doc' ament is the Decommissioning Plan (DP) for the Army Materials lechnology Laboratory (AMIL) to use in decommissioning the AMit Research Reactor located at Watertown, Massachusetts. Decommissioning as described in this DP is accomplished through decontamination. Release of the reactor building and site for uniestricted use will be requested from the fluclear Regulatory Commission (NRC) following completion of decommissioning. The
'ecation of the AMll in relation to the city of Watertown is shown in figure 1 1. The location of the reactor building (Building 100) within the r
AMit is shown in figure 1-2. 1his DP follows the format and content specified in the NRC document tit 1ed DMidRDCt 3niDhnLssion of Fe.quirrf5sJtt.s foL3Ahplitalign_19,_lgLmjmfr
.a_Rgn-ftwor Re. Actor fitq,i,li k _ Operating Licen g Revision 1, dated September 15, 1984. When the DP is approved by the licensee the DP will be submitted to the NRC for approval, and also made ave,!able to the public through local public repositories.
1.1 SUMMRY DESCRIPTION < 1he following section describes the AMIL Reactor f acility and reactor , usage during the Itcensed period, in addition, a brief discussion is given of the decommissioning alternatives considered and the cost, duration, and radiation exposure estimates for each alternative. Major tasks and schedules of the preferred alternative, quality assurance considerations, contractor involvement in the decommissioning, and the planned post-decommissioning - characterization are also summarized in this section, g 1.1.1 Reactor Facility Description The AMTL Research Reactor consists of the reactor building, certain liquid water systems rew ining inside Building 97, an underground pool water retention tank, and remaining components of the secondary coolant system. I 1-1
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i . 1.1.1.1 Building 100 Reactor Containment Building. Building 100, the , reactor containment structure (see figure 1-3). is a cylindrical pressure ! vessel 80 ft in diameter and approximately 67 ft high from ground level, with ; an elliptical top. A cross-section of the containment shell is shown in figure 1-4. The basement foundation is approximately 19 ft below ground level; a 6 ft-diameter gamma-ray facility extends an additional 9 f t below the '
' I foundation level. The perimeter walls of the gas-tight containment shell are !
2-ft-thick concrete and extend up to the crane rails 44 f t above ground level.
- The aboveground walls and the roof are covered with a 1/2 in. welded steel ;
plate. There are two large penetrations in the perimeter wall for personnel ; L airlocks permitting access to the interior of the shell. In addition, there is a large double door in the southwest portion of the containment shell. The electrical utilities are brought in through seals, which consist of a ennduit box filled with a sealant and a metal tube welded to the shell. This allows the electrical cables to pass throu2h the containment shell. The water inlet lines (city water and secondary cooling water) and outlet lines (liquid waste and secondary cooling water) pass through pipes welded to the shell. The air, l steam, return condencate, and demineralized water lines also entnr the containment shell from Building 97 through pipes welded to the steel shell.
- Air intake and exhaust are accomplished through steel ducts with flanges l welded to the shell and provided with autcmatic closing dampers. Over-pressurization protection for the shell is provided by a 2 in, line with a
- water trap equivalent to a 5-ft head of water installed between the shell and the atmosphere. The containment shell completely encloses the reactor and all of its associated equipment, with the exception of Cistnrn 242 and the secondary coolant system.
lhe containment-shell floor plans are shown in figure 1-5. The main operating floor, which is at ground level, is about 76 f t in diameter. During ,- ,
-operations, the first platform provided access to the six slant beam tubes, which exit at this level from the reactor vessel. The second platform ,
provided an area for the reactor control room and space for personnel and
- equipment for loading and unloading the reactor. A standard 10-ton crane, mounted on a circular track, was used to service the main floor, the platforms, and-part of the basement through floor openings.
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The basement area was used for experimental and operational purposes. A gamma ray exposurt experimental facility was built below the basement floor level. The gamma ray facility could also be used for fuel storage. There are 16 vertical storage tubes in one part of the basement floor that could be used for storing radioactive materials, such as beam tube plugs, collimators, and irradiated samples, lhese 16 vertical tubes are 4 ft 3 in, deep. Two tubes - 1 i are 1; in in diameter, two are 10 in in diametet, eight are 8 in, in v.iameter, and four are 4 in, in diameter. *- All the primary-ccolant equipment for the operation of the reactor is located in the basement and is separated from occupied areas by 2-ft-thick, ! ordinary concrete walls. A main sump is also located in the basement, to which all liquid waste within the containment shell drained. The liquid waste was then automatically pumped to a liquid-waste storage ta 4, which was part of the liquid-waste handling system that was located in Building 97 before its ; removal. ! The reactor is a version of the " swimming pool" type, the pool having been replaced by an octagonal open tank, completely above ground. The internal dimension 5 of the tank are 10 1/2 ft in diameter by a depth of 30 ft, l which provided 4 ft of water shielding from the fuel horizontally and 22 ft of water shielding above the centerline of the fuel vertically. The concrete t biological shield consists of an approximate inner 16 in. of ordinary concrete and 4-ft of high density concrete. Maximum neutron beam facilities were utilized by the addition of 16 horizontal 6-in. beam tubes, one horizontal
-t 6 in through tube, and six slant 6-in beam tubes. See figure 1-6 for a .
three-dimensional model of the reactor, j A cover made up of hinged metal plates closed off most of the top of the ,- pool. This allowed access to the pool for transfer of fuel elements between the reactor and the lead-lined recessed storage positions available in.tJnu , annulus. The -annulus was built around the upper portion of. the reactor pool and is accessible from the pool through a removable, watertight gate which permitted the movement of fuel elements under water. 1-8
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l P The fuel elements used in the open-top, tank-type, thermal, ; heterogeneous, H3 0-cooled and moderated reactor were Materials Testing i Reactor type assemblies fitted into a gridplate and expanded to a 7 x 9 array i of fuel elements, reflector elements, and plugs as shown in figure 1-7. The l increase in grid plate size permitted changes of core configuration and ; provided the greater fuel loading needed to overcome reactivity losses to the ' beam tubes and for operation at higher power levels. The reactor grid plate is supported on a pedestal on the bottom of the pool as shown in figure 17. - 1.1.1.2 Building 97, Reactor Facility Laboratory. The reactor facility included portions of Building 97, which provided access to Building 100 through an airlock and which also cmtained offices and laboratories for t support of the reactor operations (see Figure 1-5). The liquid-waste handling system for the reactor was also contained in the south end of this building.
'This system consisted of three aboveground 3000-gal wastewater storage tanks l and a disposable cation, anion, and mixed lon exchange system for storing and
, treating contaminated water from the reactor. Contaminated water stored in the retention tank, Cistern 242, could also be pumped to this system for processing. (According to operating history records, contaminated water was stored in the retention tank, Cistern 242, at least once.) The three storage tanks, mixed ion-exchange system, and pool fill, make up, and laboratory , domineralizer system were removed after deactivation of the reactor to make room for a particle accelerator. The portion of the piping that remains between Buildings 97 and 100 and going to Cistern 242 will be addressed in this report. Building 97 presently contains chemistry
- laboratories, an ion-inplantation facility, and a particle acceleratnr for ;
neutron production. 1.1.1.3 Cistern 242. ~A buried pool water retention tank, Cistern 242, ,- located approximately 25 ft southwest of Building 100, served as the low level waste-storage tank for the-reactor-(see Figure 1 8). The 23-1/2-ft square: .
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+ become seriously contaminated, it could have been pumped to the retention-tank. The contaminated water could then be processed in the liquid waste handling system located in Building 97. In October 1966, a_ - liquid-level-indicating recorder was-installed to facilitate waste management and to provide a method of mon'toring the retention tank for any appreciable 1eakage. The water contained in the retention tank was drained and the tank
- j. flushed to the sewer system after the reactor was deactivated.
I l T.1.1.4 Remaining Components of the Secondary Coolant System. The f 1: remaining compoaents of the secondary coolant system consist of the following i j: items; [ F
- Secondary coolant pump concrete pad I f . Three pumps j i l i
l . Secondary coolant sump beneath the concrete pad i i
. Unm und secondary coolant. piping and conduit between the sump l l area 'o the reactor building.
1 ! 1.1.2 Reactor Usage During Licensed Period i' i [ The AMTI. Reactor was used to conduct various raaterials studies, such as t-p experiments in the structure of heavy-metal azides, lattice dynamics studies L on explosive-type materials and determinations of vibrational spectra of f-- -organic secondary explosives, polycrystalline and single-crystal coherent scattering materials and liquids, activation analysis of samples containing trace inipurities, and._ inducing slight radiation effects in materials. These
- experiments were condacted during the operating life of the reactor.from June 15, 1960 to March 27, 1970.
4 i p 1-13 L l _ . - . . . - _ . _ _ , . _ . _ _ . _ _ _ _ - _- _ _ . _ , - _ , . . ~ . _ .- -
_ . _ . _ . _ _ _ _ _ . _ _ ~ _ _ . _ _ _ _ _ _ _ . _ . _ . _ _ d i 1.1.3 Comp 1rison of Considered Decommissioning Alternatives ! Following shutdown and deactivation of the AMIL Reactor, decommissioning ; planning was started. The objective of decommissioning the reactor is to obtain release of the facility and site from the NRC for unrestricted use and to terminate the license, The alternatives of safe stora0e and entombment were not considered because they did not achieve the ,Lrmy's objective of unrestricted use. The two alternatives-determined to be in accordance with , the Army's objectives are discussed in detail in the Decision Analys's Report.' These considered alternatives were partial dismantlement aad tctal dismantlement. , Partial dismantlement, as described in the Decision Analysis Report, details the removal of those interior and exterior components known to be contaminated. The remainder of the facility would be left intact for possible reuse. The total dismantlement alternative includes all work that makes up partial dismantle.4nt plus the total dismantlement and disposal of Building 100 and the reactor containment shell as clean waste. Both of these alternatives will achieve total decontamination of the facility. The. comparison of the alternatives included estimated costs, project duration, and radiation exposure to personnel; the comparisons are shown in
. Table 1 1. The basis for the. estimated cost for decontamination is given in Section 3.7. The basis for the estimated exposure to personnel is given in Section 1.4.
The Army decided upon total dismantlement as the preferred alternative for reasons explained in. the following secticn. o,av :c, this DP describes 4
- decontamination through continuous dismant'.ement as required in order to remove all contaminated materials. This will accomplish decommissioning and .'
lead :to the release of the . facility for unrestricted use. The Army will make available to the public a plan which will-describe the
. method by which the reactor building will be dismantled. This plan will be implemented once-the NRC has determined the facility meets the unrestricted use criteria.
1-14
' Table 1-1. Comparison of considered alternatives Estimated Estimated Estimated Project Radioactive Costs
- Duration d Exposure' Alternative 1$M)b($M)* (weeks 1_ (Man-remsl
- 1. Partial Dismantlement 4.3 5.1 43 10.0
- 2. Total Dismantlement 5.3. 6.1 54 10.0
- a. The estimated costs assume the radioactive waste will be disposed of af ter January 1,1992, but prior to January 1,1993. Disposal of the lowlevelradioactivpwasteduringthatperiodoftimecausesapenalty surcharge of $120/ft , which is included in each estimated cost,
- b. Thesecostestimatesarepasedonanestimatedvolumeofpackaged I radioactive waste of 8,000 ft which includes 10% void volume,
- c. Thesecostestimatesarebysedonanestimatedvolumcofpackaged radioactive waste of 12,000 ft which. includes 60% void volume.
- d. The estimated duration assumes the shipment of radioactive waste will oi *.ur in parallel with removal of contaminated components. The estimated du.ation includes both phases of the termination survey but not the evaluation of the Phase II results,
- e. The estimated radioactive exposure is the same for partial and total dismantlement because all of the exposure will occur during decontamination as part of either alternative.
l-l I s l l-15
1.1.4 Major Advantages of the Preferred Alternative The preferred decommissicaing alternative was total dismantlement. This alternative was selected because of advantages listed below:
. The building site could be returned to its original condition for ,
unrestricted use
. Total dismantlement will provide, with maximum certainty, that all radioactive contamination has been removed.
1.1.5 Major Tasks and Schedules The major decommissioning tasks to be performed by the decommissioning contractor and the estimated dur,tions are listed below. The project schedule will depend on the start date (cu rently unknown) and the number of tasks performed in parallel. A schedule 7f the entire project is presented in Section 3.8. Estimated Duration Ma.ior Tasks (weekt). A. Remove auxiliary structures 8
- 1. Cistern 242
- 2. Piping in Building 97
- 3. Secondary coolant system 9
l-16
Estimated Duration - Maior Tasks _Lwathil B. Decontaminate reactor building by removing -
-companents 12
- 1. Reactor pool internals
- 2. Reactor pool liner
- 3. Platforms
- 4. Basement piping
- 5. Basement sumps
- 6. Gamma facility and storage tubes
- 7. Reactor pool C. Dispose of radioactive waste 12 D. Perform internal decontamination 2 E. Perform Phase I termination survey 8 F.- Evaluate Phase I results 4 G. Backfill and grade 1 H. Perform Phase 11 termination survey 8 I. Evaluate Phase II results and request- 8 release of facility / site 1.1.6 Quality Assurance Plan A Quality Assurance (QA) Program shall be prepared for the decommissioning operation by the decommissioning contractor and approved by
,. the licensee. The program shall' comply with American National Standard (ANS)
QA Program requiremer.ts for iesearch reactors, N 402-1976 (ANS 15.8). l L=*.. The Quality Assurance Program shall include the following items: Review of health and safety training procedures for operating h personnel 1-17
.- Review of radiation monitoring procedures and instrumentation calibration and maintenance practices
. Review and monitor the decommissioning procedures for adequacy in regard to public health and safety, security, maintenance of as-low-as-reasonably-achievable (ALARA) conditions, choice of methods and equipment, and conformance to all applicable state and federal- regulations
. Review and comment regarding proposed changes to, or deviations from, the Decommissioning Plan
. Review of procurement documents for equipment and/or services that affect public health and safety
. Monitoring document control system with regard to work instructions and procedures, drawing and information management: radiation survey results, and field changes
. Review of documents released for NRC review / approval.
1.1.7 Contractor Participation-The Department of the Army plar.s to have the decommissioning performed by one or more outside contractors. The tasks to be performcd by contractors and the detailed performance requirements will be specified in a Statement of Work - (50W), which will become part of the contract. The contents of the S0W are described in Section 3.2.1. Contractor participation will include the following: ,-
. Providing organization and management to implement the S0W ,
. Preparing plans, including Emergency, QA, Environmental, Health, Safety, Training, Transportation, Sampling and Analysis, and Waste l Management Plans l l-18
. Preparing work procedures, as specified in the S0W c
4
. Training appropriate personnel, as required L
l . Freparing progress reports, cost and schedule reports, deviation [ .* and/or field change reports, radiation survey reports, and other i reports required by the S0W l i Conducting and supervising day-to-day decommissioning y
. Procuring services and equipment -- . Administering subcontracts and controlling subcontractors.
1.1.8 Termination Radiation Survey Plan l' .. Following completion of decontamination of the facility and removal of f' equipment surrounding the facility, as specified in this DP, a termination l radiation survey-shall be performed, following guidance in NUREG/CR-2082. I Monitorino for Comoliance with_ Decommissionino Termination Survey Critarh.2 ! Details of the survey plan are provided in Chapter 8 of this DP. i 1 The termination surveys shall include direct radiation measurements, i- analyses of smears and other samples from inside the reactor building, and l- soil analyses. i 1.2 FACILITY OPERATING AND Post-OPERATING HISTORY 1.2.1 Reactor Power History and Experiments ) ,, The AMTL Reactor was the first nuclear research reactor designed to meet
'the needs of the research programs on materials for the'U.S. Army Ordnance Corps.,-and was constructed at Watertown, Massachusetts, during the late 1950s and 1960.
i < l-19 i
-. - - . - - - . . . . ~ - - - - - - - - - - . . - - - - - _ . ~ - Initial criticality of the reactor was achieved on June 15, 1960, at a power level of 1 NW. Post-neutron tests consisting of shim rod calibrations, power calibration, temperature and void coefficients of reactivity measurements, and determinations of the worth of experimental facilities were conducted, culminating on September 16, 1960. ., Various solid-state physics research programs and experiments were ,, conducted at the 1-MW power level through June 1966 by the Army Materials and Mechanics Research Center (AMMRC). A number of local institutions (Boston College, Worcester Polytechnic Institute, University of New Hampshire, and Massachusetts Institute of Technology) also made use of the AMIL Reactor for diffraction measurements and irradiations. The reactor's license was amended in June 1966 to allow the power level to be increased to 2 MW in order to provide higher neutron fluxes for experiments. The approach to 2 MW began on June 6,1966, and was completed on June 15. The power was increased in steps of 200 kW and all parameters were observed and measured for several hours at each step. The reactor's license was updated in 1969 from 2 MW to 5 MW. On August 22, 1969, the power-escalation program began and the reactor power was increased in 1-MW steps to the maximum licensed power level of 5-MW. This program was completed with a 79-hour 5-MW run during the week of September 8, 1969, with no abnormal results observed during the power escalation. Experiments similar to those described above were planned using the higher power level, as were new experiments for advanced material and for research on, development of, and application of composite materials, improved - metal alloys, and ceramics. These experiments were performed on an irregular schedule until the reactor was permanently shutdown in March, 1970. 1-20
1.2.2 Leaks in Reactor Systems During the initial criticality, leaks of reactor cooiaat water through the concrete biological shield were observed. The leaks grew progressively worse and a first attempt to rectify the problem was to drain the annulus and apply glass tape and epoxy resin to all wall and floor joints. All surf aces of the annulus were also finished with epoxy resins. In the annulus there are cavities created by the encasement of concrete over the slant tubes as they pass through the annulus. The concrete encasements had been pro /ided with aluminum drain lines, which passed from the cavities to the main section of the annulus. The drain lines had been provided so that no water remained in the cavities when the annulus was emptied. When these lines were determined to be leaking, they were plugged, which considerably reduced the leakage in the region where the first balcony is tied to the shield. The efforts to stup the leaks met with limited success. Late in 1961, a second attempt was made to stop the leakage by drilling selected holes (2 in. diameter) and pressure-grouting a lean cement mix into the hules. This was unsuccessful because very little grout was accepted by the holes. A chemical grout (AM-9) that could be pumped as a liquid with a preset jelling time was then tried in place of the cement mix and preved to be quite successful. This n.ethod, along with the use of pressure-sensitive tape . around each beam tube joint, eliminated approximately 75% of the leaks. Some small leaks continued intermittently but did not hamper reactor operations. In 1966 a stainless-steel liner was installed in the reactor pool; this eliminated the leaks. In 1968 the volume between the stainless-steel liner and the concrete shield was connected by a drain to the basement sump in order to remove any wa'.er leakage. The water that leaked through the biological shield did not cause any major contamination spread outside the shield but is
'. suspected of contaminating approximately 50% of the high-density concrete.
The water that did reach the reactor basement drained ta the main sump. In 1961, during an unloading of the reactor core, technicians experienced difficulty in removing some cf the berylliurn oxide (Be0) reflector elements. 1-21
! An examination revealed that an element had swollen, further examination I indicated that water was leaking into the elements. Since there is no reaction between Be0 and water under reactor operating conditions, it was decided that perforated Be0 reflector elements could be used. There were no further difficulties after modifications were made to reflector elenients. , in July 1963, the heat exchanger of the reactor coolant system developed a leak. lhc heads were pulled and the corroded leaking tube was plugged. Leaks developed on four other occasions and in January 1964 the aluminum tube
~
bundle was removed and replaced. After examination of the corroded tube bundle and the well water, two actions were initiated to remedy this problem: (1) replace the aluminum bundle with one made from stainless steel and (2) install a recirculating-water cooling tower to provide secondary coolant. These actions were completed in January 1965. 1.2.3 Unscheduled Shutdowns Between January 1, 1969, and March 27, 1970, there were 62 unscheduled shutdowns of the reactor. In many instances, no direct cause was readily } apparent, as the shutdown would manifest itself as a single rod dropping without any evidence of malfunction or unsafe condition. The majority of these unexplained rod drops were helieved to be caused by noise in the period _ safety channel, which momentarily reduced the magnet current below the drop current. 1.2.4 Unplanned Release from Cistern 242 An unplanned release of radioactive liquid waste occurred from the underground liquid-waste retention tank (Cistern 242) between February 20 and February 27, 1969. The leak was detected af ter a review of records of the tank level recorder, which indicated that the level of the tank's contents dropped from 15 to 11 ft during this period. This is equivalent to a total loss of 14,000 gallons and a total activity release of 30 pCi. The Reactor Facility Safety Committee concluded that the release was well below the limit set by 10 CFR 20, Appendix C, for burial of radioactive waste in the soil. 1-22 I 1
. . __ - - . . . - .. . - - _ - . - - . . - . - - . . . ~ - - . - - - - - -
1.2.5 Reactor Operation Sununary Based on the information contained in the operations reports of the U.S. Army Materials Research Agency Nuclear ketctor facility covering the period from June 15, 1960, through March 27, 1970, and a review of the facility safety reports, there are no indications that any fuel was breached during
' ,,. reactor operations or during fuel transfers between the reactor core and the ;
annulus. Further evidence that no fuel was breached are the low levels of radioactivity and' contamination found in the reactor vessel and on the reactor internal components. 1.2.6 Post-Operating History in December 1969,-the Departmert of the Army decided to shut down the operation of the AMIL Reactor. On March 27, 1970, reactor operations were shut down and the reactor was placed in standby mode. A deactivation report was submitted to the NRC, Division of Reactor Licensing and to the Army Reactor Committee for Health and Safety (ARCHS) in December 1970. The following radioactive materials were removed from the reactor building and disposed of as follows:
. The irradiated and unirradiated fuel elements containing special nuclear material were removed under contract with National Lead Company and returned to the U.S. Atomic Energy Commission
. The beryllium oxide (Be0) reflector elements, shim-safety rods, armatures, and stainless-steel pieces from the guide tubes were disposed of as high-activity radioactive waste
'. . The fission chambers containing U-235 were transported-to another reactor facility and reported under SNM-244
. The ionization chambers were disposed of as low-level radioactive waste l-23
. The radioactive sources used for calibration and check of survey meters were transferred to the Army Radiation and Occupational Safety Branch, Army Material Command (AMC).
The water from the primary and secondary coolant systems, secondary ' coolant sump, main reactor pool. fuel storage tank (in basement), and Cistern 242 was drained and dispos?d of. Indications are that the water was monitored for radioactivity and discharged according to standard procedure, which was to either discharge to the sanitary sewer (if found to be below regulatory standards) or to dilute to achieve acceptable release ciitoria before discharging. The following liquid-waste system equipment was removed and disposed of from Building 97:
. Three each 3,000 gal liquid waste storage tanks
. Disposable ion-exchange systtm
. Pool fill, make-up, and laboratory demineralizer system
. Pumps, valves, and piping associated with the above systems.
The reactor stack, and the secondary coolant towers were removed and disposed of at a later date. The major equipment that rcmains to be removed and disposed of during decommissioning of the facility is briefly described in the following paragraph. The removal of these components is described in Chapter 3. , The stainless-steel pool liner, reactor pedestal, grid plate, and , portions of the beam tubes, control rods, and instrument rack remain in the reactor vessel. The top covers for the reactor vessel are also in place. The primary coolant system (heat exchangers, den.ineralizer systems, pumps and associated piping) remain in the reactor facility basement. The secondary 1-24
l coolant sump, three pumps, and piping to the primary coolant system were also left in place outside the containment shell. Cistern 242 and the associated piping going into Building 97 will also be removed during the decommissioning. 1.3 CURRENT RADIOLOGICAL STATUS OF FACILITY The current radiological status of the facil.tv is based on results of radiological characterization and neutron activation analysis performed. 1.3.1 Radiological Characterization The results of radiological characterization performed during 1989-and 1990 are summarized in this section. Details of the AMTL Reactor characterization, including chemical characterization, are contained in the Characterization 'leport.3 1.3.1.1 Building 100 Reactor Containment Bailding. Areas surveyed inside the building were as follows: the basement, the operations floor, the first and second platforms, the reactor control room (located on the second platform), the top of tne reactar, the ractor vessel. the reactor pedestal, and the reactor annulus. All radiation measurements were made with nort able beta-namma radictio.n instruments having detection efficiencies of approximately 10L The smears that were collected were counted at the Idaho National Eng heering Laboratory (INEL) with a decade scaler. These instruments were used for all Se areas L surveyed. All smears taken and analyzed for the reactor basement area, except smear number 34, revealed less contamination than was called for in the most restrictive part of NRC Regulatory Guide 1.86, which gives acceptable surface contamination levels. The acceptable levels for-removable contamination are: 2 less than 200 dpm/100 cm2 bata-gamma and less than 20 dpm/100 cm alpha. Smear number 34 was collected inside the storage tubes of the storage facility - 1-25 l
2 2 and read 293 dpm/100 cm beta-gamma and less than 20 dpm/100 cm alpha. Figure 1-9 shows in detail the location and number of each smear taken and contact-radiation readings (circled) in areas where readings could be detected
; in the basement.
Figures 1-10, 1-11, and 1-12 show the location and number of each smear a taken and the contact-radiation readings (circled) in areas where readings could be detected for the operating floor and the first and second platforms, respectively. Smears obtained and analyzed for these areas were all less than 2 200 dpm/100 cm 2 beta-91mma and less than 20 dpm/100 cm alpha. Figures 1-13,1-14, and 1-15 pt cvide the locations, numbers, and results of smears taken in the reactor annulus. Contact-radiation readings are also shown fer areas with the highest activities. The results of an isotopic gamma scan, performed on one of the more contaminated smears from the reactor annulus, showed europium-152 (Ee-152), Eu-154, and cobalt-60 (Co-60) in the amounts shown in Table 1-2. Figures 1-16, 1-17, and 1-18 provide the location of contact-radiation readings taken internally in the reactor vessel. Figures 1-18, 1-19, and 1-20 provide the locations, numbers, and results of smears taken in the reactor vessel. Water samples, believed to beprimary water, obtained from the reactor beam tubes contained no measurable activity. Table 1-3 provides a summary of the smear sampling, and Table 1-4 provides a sammary of the radiation survey conducted in the reactor building. From these surveys it appears that only the reactor annulus and some of the reactor components are contaminated. Transuranic isotopes were not detected on-any of the smears but, as shown in Table 1-2, Eu-152, Eu-154, and Co-60 . isotopes were detected. The highest radiation readings were ~easured on components contained within the reactor annulus and vessel. . 1.3.1.2 Building 97, Reactor Facility Laboratory. Building 97 originally contained the liquid-waste handling system for the reactor. This system was removed to make room for a particle accelerator, as discussed in 1-26
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.. Ra.sement . ,
Inside tubes of the 293 <20 storage facility- ., ,
-All other--basement 293 <20 smears Main Floor All main floor smears <200 <20 first PlatfqIm All first-platform smears <200 <20 Second P1al.fpfrm
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Annulus floor 203-725 <20 Stainless-steel piping 749-5707 <20 below reactor gate d 4 1-40
~ ' - -
_ --__.__ _ _ _ ___.___._. _ _s i Table 1-4. ;' a y cof the results of the AMTL Building 100 radiation--survey. Contact-Radiation Reading lucation (mR/h) Basement
** Demineralizer 2.0 6.0 Heat exchangers 0.3 fission product monitor 0.05 l i
Mein floor , i Californium-252 source 16.0 i Mobile N-Ray 0.2 i f_f rst Platfpng 0.2 Rractor keeper slide i Second Pletform Reactor top- 0.7 Magnets in cabinet 0.4 i Rqgsige Vessel Internals Blind flanges 50.0 550.0 Slant tubes - 8.0-30.0 Valves 10.0 60.0 < Pedestal (top) 550.0 Pedestal (bottom) 15.0 Ructor Annulus Stainless-steel racks and stainless-steel pipe 55.0' , Below reactor gate -1,300.0 i General body field by stainless-steel pipe at 3 ft 18.0 - 1-41 c + . - n.,_A.,,-.mm, ..____.mmm ,.-m. ..my__.,,,__...,,__.mm.,._..m,_._,.........,_,.__,-~__.__...__m_._, . , , , , ,#.__,m .
i i , Section 3.2. All that presently rei.iains is the piping (4-in. Iron coolant transfet line, 2-1/2 in. Iron sump pump drain lina, 2-in. aluminum demineralized water line, and 1 in. Iron test connection line) connecting Building 3 97 and 100. Smears taken from the piping inside Building 97 indicate that the radioactivity levels are less than 200 dpm/100 cm' , beta gamma and less than 20 dpm/100 cm' alpha. The piping (2-1/2 in. Iron outlet line, 4.in. iron inlet line, and 4 in. cast iron overflow line) to Cistern 242 is mostly inaccoulble since it is underground and will be surveyed during the decommissioning of the facility, lhe inlet and outlet pipes are buried 5-l/2-ft deep and the overflow pipe is-buried 3 1/2 ft deep. The overflow pipe originally went from a catch basin in Dutiding 97 into the Cistern. 1.3.1.3 Cistern 242. Cistern 242 was filled with water during the raJIological surveys conducted at AMit daring September 1989 and was not I surveyed at that time. Smears and sediment samples were also postponed until the tank is pumped. The water in the cistern was sampled during february 1990 and found to be well below the radioactivity levels of drinking water standards. There was no measurable gamma activity, other than the natural potassium 40 and radon / thoron daughters from the natural uranium and thorium decay chains. The gross alpha and beta activity levels for the sample in Ci/mL were (2 1 5) x 10"' and (1.3 1 0.17) x 10'8, respectively. Samples were not collected from the interior of Cistern 242 when f characterization was performed during the week of March 26, 1990, because the # water had not been removed. Based on the above results of.the analyr.is + of the water sample taken in February 1990 and on the low levels of ; contamination found in the reactor vessel, it is assumed that the cistern does not contain any substantial amount of radioactively contaminated materials in . the ' form of sediment or sludge. . (
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l background samples were collected in March 1990 at locations shown in l figure 1-21. These samples were analyzed for gamma-emitting nuclides by gamma . spectroscopy. Table 1-5 summarizes the results from data that were found to . 1-42 , l l l
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, give true positive. True positive results are defined as values that have a measured activity >2 measured standard deviations. Tabic l-5 includes the activity concentration with the associated statistical uncertainty, Activity Concentration (5), and the activity concentration with total uncertsinty, Activity Concentration (1). The ., statistical uncertainty includes the statistics associated with counting, backgrounds, and photopeak fitting. The total uncertainty includes ,, ~ statistical uncertainty, estimates of the uncertainty in the sample geometry (5%), and detector efficiency (5%). These uncertainties have been propagated in quadrature and are expressed as one estimated standard deviation. It is recommended that the activity concentration in the right hand column of Table 1-5, Activity Concentration (T), be used. The two samples containing positive indications of Co-60 (MTLO103 and M1LO203) were taken from the area between Building 97 and Cistern 242, where elevated radiation readings were detected during the preliminary radiological surveys performed in September 1989, The highest concentration of Cs 137 and l Co 60 detected in the samples collected from around the reactor building are ' 4.8 pC1/g and 6.3 pCi/g respectively, t The NRC has no published release criteria for radioisotopes in soil. The NRC determines whether or not a site can be released after an NRC site-specific assessment on a case-by case basis, for comparison, the Department of Energy has published release criteria for Idaho National Engineering Laboratory (INEL) soils having radioisotopic concentrations for Cs-137 and Co 60 in releaseble soil of 10 pCi/g and 4 pCi/g respectively. The DOE criteria are based on extensive pathways analysis. While the i characteristics of the two sites are diffarent, it is assumed through ' comparison that the AMIL soil concentrations have a high likelihood of being ' l iapproved for release. _However, the NRC will have' final approval _ as to whether - or r.ot the site can be released for unrestricted use. The soil samples were also-analyzed for alpha-emitters; sample number MIL l 0102 was the only sample that contain9d statistically positive Am 241 and l pu-238. The activity concentration of this sample is 0.21-pCi/g. As stated previously, the NRC has no published release criteria for soil, for l- 1-46 L m_,..._, . - - - , _ . , . ,.--,_....,_....,.__,.--..r .-m._, .. - ~ .- _._.. _ , - . . - . _ , _ . - ____ _.-_-__.-4.-.- . - - - . - - - -
compari:;on, however, the DOE INEL soil release criteria allow soil to be released with a Pu-238 concentration of 300 pCi/g and an Am 241 concentration of 80 pCi/g. It. therefore, appears that the AMTL soil has a very high probability of meeting NRC requirements for alpha-emitters, because the AMIL alpha emitting radioisotopic concentration is orders of magnitude lower than the DOE criteria, which are based on very extensive pathways analysis. 1.3.2. Neutron Activation Analysis 1.3.2.1 Introduction. During the course of operating a nuclear reactor, reactor construction materials can become activated through transformations caused by materials absorbing neutrons and possibly further transforming through various radioactivo decay schemes. In order to determine the amount of activation troducts present in reactor construction materials, it is necessary to know the neutron energies and corresponding flux intensity to which the materials were subjected, the exact nature of tha materials present, and the material exposure times. A method of verifying the activation product theoretical calculations is to obtain samples of reactor construction materials and analyre them for the concentration of activation products that are present. Due to the AMTL Reactor being nonfunctional for the past 20 years, some
-operational records have been difficult to locate. To support the decommissioning planning effort, the available data were analyzed and
. assumptions-made to facilitate an estimate of the gamma emitting activation products present in the remaining reactor construction materials. Tte only verification available is the radiological characterization discussed in Section 1.3.1. l Rnugh order of magnitude estimates of the gamma emitting activation l". L products present in the reactor support structure and biological shield were l, conducted to support the estimates of wastes to be generated and worker radiation exposures. Assumptions were made in several key parameters in the absence of reliable data so that the estimates could be quantified. In the - absence of reliable records of plant construction materials and operating history data, assumptions were made regarding the actual materials used in the reactor and biological shield, as well as the value of the flux to which the - 1-47
1 materials were exposed. Samples were not taken of reactor materials to l , support the activation calculations. These samples would have been of assistance in calculations involving the support structure and the pool liner. The number of assumptions that were necessary to allow any estimate of activation products, resulted in rough order of magnitude estimates that can be used for plannirg purposes only. Actual mecsurements of radiation fields in the reactor pool and annulus were used to form the basis for the worker exposure estimates. The presence of the unknown amounts of contamination in the biological shield as a result of leakage from the pool makes the use of activation for estimating the shield waste volumes of little value. Consequently, the waste volume estimates of concrete contained in Section 1.5 are based on assumptions of the extent of pool leakage contamination. 1.3.2.2 Materials, lhe activation calculations that were performed were based on descriptions of plant conditions contained in plant drawings from about 1965.- Not al', materials were identified on the available drawings. , . When materials were not specified, they were assumed to be of a particular type to facilitate the estimate of activation. Based on an interpretation of the plant drawing; it was estimated that there is 142,800 cm 3 of stainless 3 steel liner, 4,540 cm of carbon steel piping in the shield wall, 182,000 cm3 of carbon steel rebar,1,020 cm stainless 3 steel in slant tubes, 12,600 cm 3 stainless steel in horizontal beam tubes, 18,480 cm3 stainless steel in through tubes, and 9,439 cm 3 of stainless steel bolts. Stainles; steel materials were all assumed to be Type 316 with a density of 7.B g/cx and 3 composed of 11% nickel; 0.1% carbon; 2.5% molybdenum; and the balance iron.
-The carbon steels were assumed to be AIS!-SAE 1945 with a density of 7.9 g/cm3 and composed of 0.45% carbon, 0.75% manganese; and the balance iron. The total amounts of materials present for activation are 1.44 Mg of stainless steel and 1.47 Mg of carbon steel. .' ,
The activation of aluminum was neglected in the estimate as all .- activation products that result from aluminum have decayed in the 20 years that have elapsed since the reactor was shut down. The field survey data support this assumption. 1 48 L _ . __ _ u . _ _ _ _ _ _
! l J !
I t 1.3.2.3 Noutron flun. Data documenting the actual radial neutron flux within the reactor pool and shield were also lacking, lo allow tha activation ; estimate to proceed, a radial core flux profile wet aut'nad and core external i flux values estimated based on provided values of average neutron fluxes. In l the absence of the actual power history and fuel loadings, the activation of .
.. materials was conservatively assumed to be at saturation at the end of the i 4 operations.
An average value of fast neutron flux of 2 x 10" n/cm -set was utilized i > in the activation calculations. This value was based on 2 MW operation and } estimated radial neutron flux distribution values and pool shielding between the core and the liner. 1.3.2.4 Activt'ed Isotopes. An analysis of the gamma emitting nuclides I from neutron activation that result from the assumed construction materials is summarized in Table 1 6. Examining the number of half lives elapsed in 20 4 years indicates that the only isotopes that are present in any significant amounts after 20 years are 00-60 and Mn 54. The cobalt is present due to l production from an n.p reacticn with N1-60. The Manganese is produced by an I n p reaction with fe 54. It should be noted that although the Mn-54 half-life is relatively short, the intermediate precursor, fe-55, has a 2.73 year half life. This makes the Mn-54 contribution to the residual activity more important than its half-life alone would indicate. Thise isotopes emit gamma radiation with energies of 0.835 MeV(100%) for Mn 54 and 1.17 (100%) and 1.33 (100%) MeV for 00 60. I in addition to the gamma emitting radionuclides that are products of neutron activation, there are beta emitting radionuclides to consider. These radionuclides, such as C-14, Nb 94, and Ni-63 are not important from the standpoint of personnel exposure considerations, but may be important for the
- classification and characterization of the waste generated during decommissioning. No attempt was made to calculate the beta emitting radionu:lides because of the uncertainties previously mentioned, and because these activation components will be assessed during decommissioning activities to accurately characterize their identity and quantities.
1 49
I i Table 1 6. Principal ganna emitting nuclides from neutron activation No. of larget Activation Half-lives [ Element _ Product Utl.f-Lift ,(LQ_y.tAril Nickel Ni-65 2.5 h 70,880 Co 58m 9.1 h 19,250 i ' Co-58 70.9 h 103 Co-60m 10.5 m 1,000,000 00-60 5.3 y 3.8 Co 61 1.6 h 109,500 Co 62 13.9 m 756,000 fe 59 44.5 d 164 Manganese. Mn 56 2.6 h 67,400 , Iron Fe 59 44.5 d 164 Mn 54 312.2 d 23.4 : Mn-56 2.6 h 67.400 By utilizing the equation above, the quantities of nickel and iron atoms
- were estimated, and are shown in Table 17. .
Table 1-7. Atoms present for activation- ; , Subject Material' j isotone Stainless Steel Carbon Steel Ni-60 4.14 X-10' ? N/A 2 Fe-54 6.37 X 10' i 9.12 X 10' 1. 1.3.2.5 Calculations. - The number of target atoms for. each isotope of - .
-concern was estimated using the equation -
.Nn =(f) (g) (a) (A) (aw) i.
l l_ l-50
, _ .' .m .._...~_-~..,..J. ,_.._.O_..... .,_-..,.-......,~..i.....,_.s_,_,r..,_.,-,.._m,-,....__., ..m,.__.._m.,.m.,,,, . ,
where No = the number of target atoms present at time (0) f - the fractional amount of element present 9 = the total gram quantity of the parent material a = the atom percent of the target radionuclide
~
A - Avogadro's number aw - gram atomic weight of the element of concern Given the number of target atoms and estimating the fast neutron flux present for activation, the activity of the resulting radionuclide can be determined by using the following equation. A-(o) (() (No ) (c'") where A - Curie amount of the product radionuclide o - reaction cross-section d e fast neutron flux ho number of target atoms present at time (0) e'" - decay factor, t equals 20 yrs Performing this calculation for the two isotopes of concern resulted in an estimate of 23 Ci of Co-60 and 0.26 mci of Mn-58, These activities would result in stainless steel having an approximate average specific activity of 33 pCi/g of Co-60 and 76 pCi/g of Mn-54 and carbon steel with 102 pCi/g of Mn-54, 1.3.2.6 Conclusion, it should be noted that many assumptions were made to arrive at virtually all of the key input factors to the activation calculations, This has resulted in the values listed above being rough estimates. The values presented here are felt to be conservatively higher than are expected to be encountered during decommissioning, Samples of the wastes will be analyzed as they are generated for the purpose of waste classification. 1-51 1
1.4 [STIMATED PERSONNEL DOSE The estimated dose to workers during decommissioning activities is calculated from estimated task duration, estimated crew stic, and estimated average radiation field. The estimated task durat,un is the duration of the task during which workers would be exposed to rsdiological hazards. The , estimated crew size only includes personnel working in the estimated field. The estimated average radiation field is based on measured radiation fields. . The estimated radiation dose for each Lask required to perfor.n decommissioning of the AMIL Reacter is given in Table 1-8. The sum of the estimated exposure for all the tasks is 10.0 man rems as shown in Table 1-8. The actual radiation dose to personnel, however, will be kept to levels ALARA by utilizing engineering controls during decommissioning activities. 1.5 ESTIMATEo VotuME or RADIOACTIVE MATERIAL TO BE REMOVEo
-The estimated volume of radioactive waste generated during '
decommissioning of the AMIL Reactor is summarized in this section. The waste volume estimate is based on the following assumptions:
. The secondary coolant sump and associated piping is not contaminated.
. The waste holdup tank beneath the reactor is empty but contaminated.
. The concrete shield is 50% radioactive waste due to contamination-from pool leakage. (The activation of the concrete is insignificant ,
compared to the assumed contamination due to pool leakage).
. Cistern 242 is uncontaminated. -
1-52 >
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, tt r .
aac l a mi( t . id o t a T sR E r o t c de a ez . ti e aS . R .m 3 4 4 6 0 0 0 0 _ 1 1 1 1 L i w T t e M sr A EC e h t f o ) .- g r _ n h i i n k n 6 o so 2 5 0 4 4 5 2 8 i ai 3 2 3 8 1 3 8 s Tt s a i m r m u o D c e D r o s d t n f u e l l t e d e d e e u a e d l t l e e m s n i t n e e o n c s a i r i s D a o - h c h a s s , s n s b n r s s g o o o a e n e c t e n i t l i t i t n c d n p e y e a o a n i i r t r e _. i i e a a p c i c s d t p r t n s n a a l s , o n o b R i r m a s r' c e c . o t l d l d c r s e e' r r e t e a
. e
.D s f s ep d'
e s s e g a n e h n g n i t e s u e s _ m k t n v .h c' i h o v i r _ 7 t s a e r r xs r r r s T n o o er o t o o o E o t' t e t t t p c' c tz c c c c m a a ,ai a a a a _ . o e' e el e e e e 8 c re rr .ha r r r r r 1 e ea e ee e ed e e v vw vr vn oiv o oa vi v v o e o od oe o. . l b m e ea mr mn ei mm m ee e < mh' J m a a R Rh Rl Rd R 2s 6 . R T [w . ii!!!; l;ij:j , -i;;! ij ;! ;4 , b ,5S! !! *
..___m _ _ ___._ _ _ _. __ _ _- - -
in the atuence of concrete coring data, the amount of concrete shieid F contamination can only be estimated because of unknown amount of contamination as a result of pool leakage. The concrete shleid material will be surveyed . during removal of the concrete shield. Only material meeting NRC release criteria will be disposed of as uncontaminated waste. All other waste will be
, disposed o' as radioactive waste after it is analyzed to identify radioisotopes and their specific activity. '
Activated material quantities were estimated as explained in Section 1.3.2, Analyses of radioactive waste will accurately characterize the i waste to identify gamma emitting and beta emitting radionuclides that are l products of neutron activation. The estimated contamination waste volume, activity, and principal radionuclides expected are shown in lable 1-9. The potentially contaminated lead listed in lable 1-9 is from the lining of the floor storage pit'., boxes in the annulus, and sleeves around the outer portion of the beam-tubes. This potentially contaminated lead will be removed I and surveyed. If contaminated with fixed contamination, it will be packaged and given to the l kensee for storage as mixed waste. If it is contaminated with removable contamination it will be decontaminated and recycled. If the j lead is uncontaminated, it will be recycled. Tablo 1-9. Estimated radioactive waste { EstimatejVolume Principal Malerial ift i _ Activity _. RidliLnacitdch)" l Stainless steel 130 23 Ci 00-60= l Steel 200 0.15 milli Ci Mn 54 b Lead 51 unknown unknown l Concrete 6.750 unknown Cs-137, Sr-90, Co-60
- a. Beta emitting-radionuclides produced by neutron activation are not ,
included. These activation components will be identified by analysis-of - > waste during decommissioning,
- b. If the lead is contaminated with fixed contamination, it will be packaged and given to the licensee for storage as mixed waste.
l-54 s I
1.6 DECOMf41SSIONING ALi tRN ATIVE lhe proposed deconar.issioning alternative, decontamination, will achieve the decommissioning objectivc of releasing the site for unrestricted use and termination of the license. 1.7 DEc0MMISS10NINc ORGANIZATION AND RESPONSIBILITIES The decommissioning operations of the AMIL Reactor facility will be performed by a qualified contractor under a contract awarded and adm' istered by the U.S. Army. A qualified decommissioning contractor shall have decommissioning experience with a facility similar to the AMIL Reactor facility. Decommissioning requirement s are generall ' specified in this Decommissioning Plan. Specific requirements for the decommissioning contractor, including saf ety requirements, will be specified in a Statement of Work to be finalized af ter approval of this DP. The statement of Work (50W) will include appropriate parts of this DP and will become the major specification for the AMil Reactor decc .missioning operations, lhe appropriate parts of the DP to be included in the 50W are all parts constituting decommissioning requirements or information useful to prospective contractors during preparation of their proposals. However, some information, such as estimated cost, would not be included in the 50W. The AMIL Reactor decommissioning project organization relative to safety is shown in figure 1-22. A brief description of the key positions in figure 1 22 and the responsibility of each are given below. As the licensee, the MIL Commander is responsible for the cverall decommissioning project and has authority in all associated mattcrs, including project safety. His representative for the decommissioning project, including safety, is the Project lechnical Monitor, an MIL employee. He has the responsibility and authority to monitor the decommissioning of the reactor. The MIL Technical Monitor will have direct access to the MIL Commandor in order to recommensi work be stopped. 1 55
.t i
i
!NRC. '
I t j l ! r 3
- MTL
! (Licensee) Radiation , MTL Commander ! ControlCommittee [ ! 1 - 1 t
- New England Division MTL Contracting Officer Technical Monitor. l l Radiation Protection Officer ' !
f S i i independent Contracting Officer's f Representative
/ Quality Assurance '
- Evaluator /HP Contractor j
Project Manager Radiation Controland Safety Ofreer ; l Health Physicist ! l Certified Hygienist ( Quality Assurance Evaksator f i f
?
i T91CSO3 I Figure 1-22. AMTL Reattor decommissioning organization relative to safety. i l t l l I i !
The Corps of Engineers (005.) New ingland Division (NID) shall act as Mll's Contratting Of ficer. Mll's Contracting Of ficer (an legally stop the work based on persanal knowledge or on input f rom t he Projec t 19tbniral
- 1 Monitot, if appropriate. The Cont racting Of fiter's Represent ative (COR) will have full authori'y of administering the decommis:,ioning contract. The CDR shall also be providrd by the Col, Niu.
An Independent Quality Assurance Ivaluator/ Health Physicist (oaf /HP) is responsible for the continuous monitorir,g of all decommissioning activities and will report regularly to the L icentee, Project Technical Monitor, and the Contruting Officer (KO). The QAl/HP will be experienced in decon,missioning projects and will ensuie contractor compliance with all provisions of the contract including sr.fety requirements, lhe Radiation Protection Of ficer (Rp0), an MIL employee, is responsible f or reporting any safety violations and te:hnical inconsistencies with installation procedures. ihe RPO has the authority to directly stop work on any operation which he has determined te be in violation with 10 CIR and the licensing order for decommissioning. The AMIL Radution Control Committee (RCC) will advise the licensee, as needed, in matters related to radiation safety during decommissioning. Organization and conduct of the RCC shall be in compliance with Aru,y Reaulations 385 11, lonizing Radiation Protection, lhe decommissioning contractors' Project Manager is the person assigned to the project by the contractor, lhe Project Manager will be experienced in decommissioning projects and will have coniractor responsibility f or all aspects of the project including safety. The contractor's Radiological Control and Safety Of ficer (RC&S0) will report to the Project Manager and will have responsibility for safety during decomniissioning operations. The RC&SO will be a health physicist and will al:,o have received extensive training in industrial saf ety and industrial " bygiene. Reporting to the RC&SO will be a health physicist and a certified industrial hygienist (ClH). lither the Cid or the RC&SO will be an OSHA certified safety professional, 1-57
l r i I The contractor shal) also provide a 0AE to monitor the project and ensure contractor corpliance with all provisions of the contract including safety, i 1.8 REGUL.ATIONS, REGUL.AYORY GUIDES, ANI. STANDARDS The terms " regulation," "guldeline," " standard," and
- criteria" are often used interchangeably, but there are distinctions. Regulations are rules .,
having the force of law and are issued by an executive authority or a government. A guideline is a recommended practice or guiding information supplied by an agent with implied intimate technical knowledge. A standard is established by " authority" as a rule to follow. In general, standards set forth limits or definitive ways of accomplishing an objective, whereas criteria provide a yardstick for comparison as a basis for judging the acceptability of a practice. I This section identifies and discusses the regulations, guides, and ! standards applicable to decosnissioning of the AMTL Reactor. 1 1.8.1 Applicable Regulations federal regulations that are applicable to decommissioning research reactors appear in-the Code of federal Regulations (CFR). While all the fMers' government regulations are contained in the CFR, different titles are atsociated with various government agencies, commissions, and administrations. For example Title 10 -Energy, pertains to the NRC; Title 29- Labor, includes eorkes health and safety; Title 40--protection of Environment, includes regulations of t w Environmental Protection Agency (EPA); and Title 49--Iransportation, deals with transportation of hazardous materials. , Some of the regulations under these titles have immediate applications-in decommissioning..and some have application by implication of related subject matters. 1.8.1.1 Code of Massachusetts Regulations (CHR), CHR- Title 310 Hassachusetts Air Pollution Control Regulations 1-58 l
1 i J CMR . Title 310, Chapter 30 Massachusetts Hazardous Waste Management ! i Rule (MR Title 310, Chapter 40 Massachusetts Oil and Hazardous Release 3 Regulations i: CMR Title 310, Chapter 7.10 Noise. 1.8.1.2 Code of Federal Regulations. 10 CFR Pert 19 Notices, Instructions, and Reports to Workerst Inspections 10 CFR Part 20 Standards for Protection Against Radiation 10 CFR Part 30 Rules of General Applicability to Domestic Licensing of Dyproduct Material- : 10 CFR Part 50 Domestic licensing of Production and Utilization facilities (Note: The termination of the reactor license must be in accordance with 10 CFR 50) 10 CFR Part 51 Licensing and Regulatory Policy and Procedures : for Environmental Protection 10 CFR Part 61 Licensing Requirements for Land Disposal of RLdioactive Waste , 10 CFR Part 71 Packaging of Radioactive Material for Transport and Transportation of Radioactive Material under ; Certain Conditions i 10 CFR Part 140 financial Protection Requirements and Indemnity Agreements 29 CfR Part 1910 Occupational Safety and Health Standards i 29 CFR Part 19?6 Safety and Health Regulation for Lonstruction 40 CFR Part 260 Hazardous Waste Management System General l , 40 CFR Part 261 Identification and Listing of Hazardous Wastes Standards Applicable to Transporters of
., 40 CFR Part 262 r Hazardous Waste 40 CFR Part 61 National Emission Standards for Hazardous Air Pollutants 40 CFR Part 141 National Primary Drinking Water Regulations 1-59
_.- . ~ _ _ . , . . __ _ . _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ . . _ _ _ _ _ . . - . ~ . . _ . . -
49 CfR Parts Department of Transportation Hazardous Material 110-199 Regulatlons. 1.8.2 Regulatory Guides in addition to regulations that carry the force of law, regulatory , bodies such as the NRC and IPA prepare regulatory guides that, among other things, suggest agancy approved methodology and soluttons to problems. lhey ,, generally provide the most ef fective method of obtaining approval for a particular course of action. The NRC Regulatory Guides applicable to this project are as follows:
' NRC Regulatory Guide-h!n3her _.__ __ _U.t.l.e _ _ _ _ .
1.8 Personnel Qualification and Training 1.16 Reporting of Operating information 1.86 lermination of Operating Licenses for Nuclear Reactors 1.143 Design Guidancc for Radioactive Waste Management Systems, Structures, and Components Installed in Light Water Cooled Nuclear Power Plants ' 3.X Draft Standard format and Content of Decommissioning Plans : for 10 Cf4 30, 40, and 70 Licenses. l 8.2 Guide for Administrative Practices in Radiation Monitoring- ! 8.3 film Badge Performance Criteria 8.4 Direct-Reading and Indirect-Reading Pocket Dosimeters 8.6 Standard lest Procedures for Geiger-Muller Counters 8.7 Occupational Radiation Exposure Records Systems 8.8 Information Relevant to Ensuring that'0ccupational Radiation ,- I Exposures at Nuclear Power Stations Will De As low As Reasonably Achievable r 8,9 Acceptable Concepts, Models, Equations, and Assumr 'nns for ; a Bioassay Program 8.10 Operating Philosophy for Maintaining Occupational Radiation Exposure As low As Reasonably Achievable 1-60
8.15 Acceptable programs for Respiratory Prot 9ction N/A Guidelines for Decontamination of facilities and Equipment Prior to Release for Unrestricted Use or lermination of ticenses for Byproduct source, or Special Nuclear Material,
.- August 1987.
1.8.3 Standards A number of institutiona or technical societies, such as the American Society for Testing and Materials (ASTM), the international Comah,sion on Radiological Protection (ICRP), the Natinnal Committee on Radiation Protectics. and Measurement (NCRP), and the American National Standards institute (ANSI), publish standards. While not carrying the force of law, they do represent the formal statement of inhnical opinion of the bodies issuing them. AUSJ_and ASTM StJMardi Af4SI N13.13 Control of Radioatiive Su; f ace Ccnt am; nation of Material Equipment, and Iacilities to be . Released for Uncontrolled Use (Oraft) ANS! /88.? 1980 Practices for Respiratory Protection ANSI N13.1 Guide to Sampling Airborne Radioactive Materials 7 in Nuclear f acilities ANSI N323-1977 Radiation Protection Instrumentation Test and Calibration ANSI /ANS IS.10 1981 Decommissioning of Research Reactors ASIM E 1281 Standard Guide for Nucleas Facility Deconuissioning Plans. 1,8,4 Applicable Army Requirements and Regulations U.S. Army Corps of Engineers Rafety and Health Requirements Manual, EM 385 1-1 Army Regulation No. 385-11 (AR 385-11), Innii ing Radiation Protection 1-61
Army Regulation No. 4014 (AR 4014), Medical Services Control and Recording Procedures for Exposure to lonizing Radiation and Radioactive Materials. 1.8.5 Informal Guidance and Technical Reports , Informal guidelines published by the NRC can be found in NUREG ,, documents, Bratch Technical Position papers, inspection and Enforcement Branch notices, and other external or internal documents. There are numerous technical reports published by the NRC and the DOE that support the subject of deca missioning research reactors. The following NRC documents are directly applicable:
- NUREG/CR 1756 " Technology, Safety, and Costs of Decommissioning Reference Nuclear Research and Test Reactors" and addenda
+ NUREG/CR 2082 "Honitoring for Compliance with Decommissioning Termination Survey Criteria"
+ NUREG 0586 " Draft Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities"
- NRC's " Guidance and Discussion of Requirements for an Application to Tarminate a Non power Reacter facility Operatinn License"
+ NVREG/CR-2241 " Technology and Cost of Termination Surveys ,
Associated with Decommissioning of Nuclear facilities."
~.8.6 Permits / Licenses Covering the AMTL Reactor.
- snel;y [icense/Pelmit Numbfr NRC facility Operating License R65 1-62
i 1.9 TRAINING AND QUAL rtcATroNS The following is a summary of the training program. Details are discussed in Chapter 2. 1he training and qualifications of personnel will depend on the
.. individual task assignments and the experience of the personnel assigned to the decommissioning activities of the AMTL Reactor. Training and qualifications of personnel shall comply with ANSI 3.1. Only qualified personnel will be assigned to the AMIL Reactor decommissioning activities.
1.9.1 Training Program Descriptions Training topics will depend upon:
- 1he health and environmental impacts of planned operations
- Applicable regulations, standards, and guidelines pertinent to operations involving radiologically or chemically hazardous materials / waste
- The purpose of the training
+ The personnel to be trained (e.g., their education, training, and experience).
Documentation of training shall be by appropriate Environmental Health
. and Safety (EH&S) form, currently " Training Record Sheet."
Personnel having received substantial radiation safety training within the past year may, upon demonstration of their knowledge to the satisfaction of the Project Health Physicist,.be exempt from general employee training. 1-63 l l-
t L 1he anticipated training programs include: i l l
+ llatardous Waste Operations Training: The contractor shall comply ;
with the safety and health requirements for hazardous waste operation as prescribed in 29 CFR 1910.120. In addition, ., ; alternate workplace standards recommended in publications related to workplace exposure criteria, such as the Threshold Limit Values ,, and Biological fxpostre Indices by the American Conference of Government Industrial flygienist, shall be used in lieu of OSilA l standards, where OSilA standards are less stringent or do not exist. '=
- General imployee Training: General employee training in compliance with Title 10 CfR Part 19.12 will be required for all personnel involved with radioactive materials or working in the vicinity of radioactive materials.
+ Respiratory protectiti: Respiratory protection training will be implemented to meet project requirements in cepliance with ANSI Z 88.2, NRC Reg Guide B.15, NRC HUREG-0041
- Manual of Respiratory Protection" and 29 C}R 1910,134.
+
liearing Conservation: A hearing conservation training program will be conducted to implement 29 CfR 1910.95.
- Hazard Communications: llazard communications training in compliance with 29 CFR 1910.1200 will be conducted as applicable.
Technical Training: Job activity simulations or brtefings shall be conducted daily to address proper handling and use of equipment, health and safety issues, and ALARA considerations. ' These will depend upon the task and personnel and will be documented in the training record P
^
L 1-64
1.9.2 Administration and Recordkeeping The Project Health Physicist will be responsible for the training program ana maintenance of personnel training, qualification, and exposure
, records. The Contractor's responsibilities for training are specified in Section 2.1.2 and 2.2.2.
Complete up to date training, qualification, and exposure records will be maintained oa all personnel. 1he records will include: ,1 N
+ Bioassay analysis e Personnel exposure records
- Individual dosimeter readings as related to daily tasks and work proceoures
- Respiratory protection qualifications (medical clearance and fit test)
- Audiogram results n
- Training records
- Visitor logs and exposure information L
9 l-65
1 r CHAPTER 2 OCCUPATIONAL AND RADIATION PROTECTION PROGRAMS ,
- 2. INTRODUCTION lhe Occupational and Radiation Protection programs (ORPPs) for the AMil
" rer tor decommissioning project consist of a set of policies, procedures, and instructions to protect workers, the general public, and the environment.
Objetlives of the ORPPs include:
- Ensuring the health and safety of personnel by providing protection programs that include a commitment to the principles of maintaining exposures to ALARA levels
- Minimizing the exposure of the general public and the environment to the radioactive and/or hazardous chemical ef fluents that may be ,
released during deconmissioning activities
+ Identifying and separating contaminated structures, Surfaces, systems, and components from those which are not contaminated
. Disposing of contaminated and noncontaminated components and materials properly and safely
. Ensuring that the facility and site meets all radiological decommissioning requirements and is ready to be released for unrestricted use.
The ORPPs_ provide integrated occupational. health, health physics,
. industrial hygiene, and safety elements. 10 meet NRC n; porting guidance, these elements are discussed in Sections 2.1, Radiation Protection Program, 1 and 2.2, industrial Safety and Hygiene Program. This format creates
-repetition in the text because normally a respiratory protection program is provided for all airborne hazards for both radiologically and chemicaG !
2-1
,_ _ . . . _ _ . , . .. ..__-.____.___.____.__.<__...u. . __
l 4 4
)
i hazardous substances, not in separate programs. Additionally, personnel training is an integral f acet of both portions of the ORPPs. and training is ) discussed in Section 2.2.2. However, a brief discussion of certain training : items was presented in Section 1.9, Iraining and Qualifications. ! 2.1 RADIATION PROTECTION PROGRAM , l The Radiation Protection Program (RPP) for the decommissioning project includes requirements to monitor radiation and radioactive materials, to control distribution and releases of radioactive materials, and to keep
^
radiation exposure for individuals and the collective radiation exposure i within the limits of 10 CFR 20, 29 CFR 570.57, and at ALARA levels. . i 2.1.1 Personnel-The selection, training, and qualification of deconunissioning personnel shall comply with the criteria specified in ANS 3.1. The decommissioning contractor's personnel will become f amiliar with the location and magnitude of sources of radiation to which personnel may be-exposed during the course of work. In addition, health physics personnel will become familiar with the use of approved Radiation / Hazardous Work' Permits (RHWPs), Standard Work Permits, and detailed Work Procedures. The deconunissioning contractor's health physics personnel will include: Radiological Control and safety Of ficer (RC&S0) l .
+
Radiological Engineer / Health Physics Supervisor (RE/HPS)-
+ _ Health physics technicians.
ORPPs procedures will be prepared by the decommissioning contractor in , accordance with ANSI 15.10. The licensee will require they be reviewed by the 2-2
AMll Radiation protection Officer and other appropriate AMit safety personnel, ! Implementation guidance will be provided by the RC&SO. 2.1.2 Notices, Instructions, and Reports to Workers, inspections Notices, instructions, and reports shall be given to individuals '
" participating in the decommissioning of the AMTL Reactor. These notices, ,
instructions, and reports shall be in compliance with 10 CFR 19. [ In addition, NRC inspectors may consult private 1" with decommissioning workers during inspections. The conduct of the i ,ections and the rights of the decommissioning workers relative to radiological working conditions are . f specified in 10 CFR 19 and shall be communicated to decommissioning workers. 2.1.3 Training Although all persons will receive tioining, not all persons should receive the same type of training. To prevent duplication and to make efficient use of time, project personnel will be grouped into three categories and will be given training commensurate with potential radiological problems to be encountered-in their scope of work. The three groups formed are:
+ l'onradiation workers
- Radiation workers directly involved in handling i radioactive /centaminated materials and entering radiation areas
- Radiation protection technicians, t The ORPPs shall include a training program for all persons who will be involved in the decommissioning project. All decommissioning personnel shall receive instruction concerning radiation protection through p orientation / training. Each worker will attend one of these orientations and will be evaluated by examination upon the conclusion of the training. A
. passing score is required before these personnel may be occupationally exposed 2-3 eve rw 7'wm-=9e1-=#,ew e<e h- 1 e rer- r-tw-- mtw m:=-we--- t e-er e em vvt-=N & ew re w1- 9-1rtv ve et-ww sii -ur'ew-i e r- re- + t -= v wvt ur- ~wv- -N** W W 'stt T-T"v '-IF**u*V'*-- Y ' pr Y m w y'i-sy' a'y-yyyw-
. . - . . - . ~ . . - - . - .- .- - . - . . - . - - . . . - - - - - -
to radiation 'The training program will adhere to the guidelines of the Institute for-Nuclear Power Operations (INP0) General Employee Training "rogram. The RC&SO, or a designee, shall adn.inister comorehensive radiation worker ' tr.,ining, OSHA hazardous communication iraining, at.J su indoctrination session for all worh rs and to all on site management personnel. Additional briefings - and practical-factors training shall be performed on 1 routine basis to -; familiarize personnel with work procedures, equipment, radiation control requirements, and hazards associated with the various work elements. Documentation of individual training and qualifications shall be filed and maintained in the 5ealth physics office at the work site. Records of training shall be maintained: these will include trainee's name, training date, subjacts covered during training, equipment in which training was received, results of written tests, and the instructor's name, Specialized training shall be provided to employees before they are '
' allowed to undertake jobs with high exposure potential, Employees and supervisors shall receive this training as part of the crew. : Mockups and other training aids may be used to traid the workers so that time spent in high-radiation fields is minimind.
- Objectives of the trif 1A arogram are to accomplish the following:
+- Provide involved personnel with information about radiologically ar.d chemically hazardous substances, sources and types, exposure routes, and effects Provide inform & tion on the ORPPs for the decomm ioning project in order to enable each person to comply with health and safety rules .
ar.d to respond properly to all conditions
+- Provide W*ruction in the fundamentals of radiation and chemical protection to enable individuals to maintain their-own exposure and colle.tive exposure ALARA 2-4 i
. --.. . - - - ... - -.-.~.~ -
L -. Provide information and training on personal protection equipment, monitoring instruments, and equipment availabic, and how to use them l + inform each person abott- NRC, EPA, Occupational Safety and Health Administration (OSHA), state and local regulations and requirements, and other applicable rules and regulations concerning health and safety. , 2.1.4 Adninistrative and Radiological Controls Administrative and radiological controls ccmprise the measures taken to hmit r6 stion exposure to personnel and the spread of contamination. I 2.1.4.1 Exposure Limits. Limits on the radiation exposure of individual workers involved in radiation-related work have been set for the nuclear industry by the NRC and are applicable here. These limits are stated in u 10 CFR 20, " Standards for Protection Against Radiation." However, in order to ensure that individual and collective doses are kept ALARA, the contractor shall establish work procedures and a radiation / hazardous work permit system . l' to ensure that all work performed is evaluated with respect to the ALARA L philosophy
- ring the decommissioning of the AMTL reactor.
l Per:onnel 18 years of age or older classified as radiation workers shall l' have their whcle body doses administratively controlled according to the guidelines listed in Table 2-1. Occupation whole body dose limits may be I permitted to exc(cd Table 2-1 administrative guidelines, provided tha't an L approval has been signed by the-RC&SO. Under no circumstances will the limits in 10 CFR 20, listed in Table 2-2, be e>cceded. 7 Radiation exposure limits to any individual who is ander the age of 18 i
. years are specified in 10 CFR 20.104.
Visitors and ronradiation workers will generally not have access to the radiation control area (RCA). If access is required, exposure to radiation 2-5 l l
l l Table 2-1. Administrative guidelines for radiation whole-body doses during decommissioning P Administrative Guidelines : (mrem) Nonradiation Workers
- and Visitors Radiation Workerg Hourly 0,2 --
Daily 2 100 Weekly 10 300 Calendar quarter 13 500 Calendar. year 50 1,000
' Table 2-2. Regulatory limits for radiation doses during decommissioning for a calendar quarter (mrem)
-- ==
Regulatory Limits for Radiation Workers' _Lmrem) L Whole body, gonads, 1,250 g blood-forming organs l Lens of eye V . L Hands and forearms, 18,750 - l feet and' ankles 1 l Skin of the whole body 7,500 . i
- a. In accordance with an agreement'between the Army Material Command (AMC) and the NRC, 10 CFR-20 requirements will be kept until January, 1993.
2-6
.-- . . . . . . . . . , , _ . - , . - . _ - - . - . - - _ . _ - _ . _ - . _ , - _ . . . . - _ . . . . - _ . - - , . ~ . _ , , , , - . . .
1 will be kept below Table 2.1 values. Visitors and nonradiation workers who are allowed access to the RCA shall be escorted by a radiation worker, ;
-according to requirements specified in standard procedures, whenever they enter the RCA. These nonradiation workers must be required to wear radiation dosimetry and will have exposure levels documented.
4 The dosimeter type and frequency of exchange shall be specified in procedures to be prepared by the decommissioning contractor and approved by the AMIL RPO. A direct reading radiation dostmeter will be worn in conjunction with personal dosimeter badges in high radiation areas. All dosimetry shall be National Voluntary Accreditation Program (NVLAP) accredited. To ensure compliance with ALARA principles, the contractor's RE/HPS will be available to review the work permits and assist in preparation of work permits. An ALARA checklist shall be prepared by the decomniissioning
- contractor and approved by the AMTL RPO. The ' checklist shall be used to ensure that all work is preplanned to minimize radiation exposure. The checklist shall include the physical and administrative implcmentation of the-radiation exposure controls.
The RC&SO. implements the ALARA philosophy. He/she also reviews and
! approves Lprocedures _concerning work that has 'the potential for occupational exposure. Appropriate health physics procedures shall _be referenced in the Work Procedures' to ensure that any occupational exposure is maintained ALARA.
Entrance to the restricted areas of the facility shall be controlled by the RE/HPS and requires the issuance and' approval of Radiation / Hazardous Work Permits (RHWPs). The description of the RHWP system is discussed below. -When
. - the Permit is -initiated, =the work assignicent and applicable procedures are developed and listed. As determined on a case-by-case basis, additional-
. health physics procedures can be implemented at that time, if needed.
Public radiation exposure resulting from decommissioning the AMTL reactor must comply with 10 CFR 20. The maximum public exposure limits for external exposure are specified in 10 CFR 20.105, " Permissible levels of Radiation in 2-7 l a. l
Unrestricted Areas." Limits for internal exposure pathways are given in 10 Cf R 20.106, " Radioactivity in Ef fluents to Unrestricted Areas." As in the case of occupational exposure, 10 CFR 20.l(c) requires application of the ALARA principle to the control of public radiation expotures and releases of radioactive materials to the environment. During the decommissioning of the AMIL Reactor, no measurable public radiation exposures or releases of , radioactive materials to the environment are expected because of engineering controls during decommission. However, an environmental monitoring program , shall be in place, and it will be in accordance with ANSI 15.10 to demonstrate compliance with 10 CFR 20. 2.1.4.2 Radiation / Hazardous Work Permits. RHWPs shall be established to ensure that hazardous conditions and protection measures are identified and communicated to those who will perform work in potentially hazardous areas or who may work with material that may be radioactive or radiologically contaminated. The procedures snall also provide the mechanism for exposure accountability and ALARA assessment. A RHWP shall be issued for a specific task, job, or series of tasks to be performed within restricted areas; with material which is radioactive, radioactively contaminated, or chemically hazardous; or with high-hazard operation (confined space, platforms, crane operation,etc.). RHWPs shall be used for areas where hazards are significant and may change. RHWPs shall establish:
+ Mapping radiation and contamination based upon radiological surveys, analytical results and calculations
. Segregating the available work area into sections (e.g.,
contaminated, clean, working area, examination area)
- Limiting access to control the spread of contamination from -
contaminated to clean areas and limiting access to all personnel who are not directly involved in the specific task
- Describing the methods to identify and mark all removed items, and noting their place of origin and any other pertinent radiological information 2-8
. = . - . -. - .-.- - - - - - . ..- -..
n ;
. Packaging contaminated wastes in appropriate containers (as r prescribed by NRC and 00T regulations and radwaste disposal site ;
criteria)
. Maintaining accurate shipping records throughout the operation I
- Listing work area monitoring requirements, which warn of any I unexpected changes in the radiological conditions
. Listing personnel monitoring and protective devices I
. Listing requirements for maintaining accurate and updated records of personnel exposure, surveys, and lessons learned in order to improve !
-and revise procedures as necessary. l l~
i- The superviso of workers who perform work under RPWPs shall be responsible for ensuring that the workers have been properly prepared prior-to entry into a restricted area. Proper preparation includes: n
. _Successfully completing Radiation Worker Training and OSHA Hazards Communication Training and Respiratory Protection Training (if-l applicable) i . Checking workers' dose records to ensure entry and/or work without- >
L exceeding established limits (administrative and regulatory) Providing pre-job ALARA briefings,- training, or instruction, if
~
recommended or required 1 L
- Ensuring that approved detailed procedures covering the total 1
radioactive work aspects have been prepared prior to the start of work
. Ensuring that appropriate procedures, tools, and equipment-are available to perform the job 29
Ascertaining that all workers and supervisors have read and understand the RHWP and its requirements, as well as the work conditions and special controls necessary. , The RC&SO shall Le responsible for ensuring that personnel have been , properly prepared for entry before approving their entry to the restricted area. The RE/HPS, the work superviso. , and the workers responsible for performing-the work shall : ensure that all radiation / hazard controls are properly implemented throughout the job cycle. RHWPs will- be valid only for the period of the task (s) to be performed and only for the specific task (s) indicated on the RHWPs. RHWPs are provided
'for entry and-work in areas where radiological conditions are subject to ,
significant or unexpected change;. therefore, additional instructions or requirements shall be incorporated in the RHWPs, as changes warrant, through the following procedures: Any supervisor responsible for completion of a task to be performed within'the RCA requiring a RHWP may request one by completing the applicable portions of the Radiation Hazardous Work Permit Request Form and forwarding it to the RC&SO and RE/HPS, Requests for RHWPs should be submitted a minimum of 24 hours prior to scheduled job initiation. The RC&S0 and the RE/HPS will. review the request and prepare the RHWP, 6fter determining the following: The radiological status of the work area, through the appropriate contractor monitoring and surveys. - The necessary precautions, based on radiological status, . including protective equipment, special control measures, and actions to reduce exposure to ALARA levels. 2-10 i
+ .--- -- , , , .
l
. Each RHWP must be assigned a number indicating the year and month of issuance and the order in which the RHWP was issued that month. .
Example: The first RHWP issued in November 1991,.would be numbered 91-11-01.
- All RHWPs shall be reviewed and approved by the RC&SO and RE/HPS.
The work supervisor will sign all RHWPs, indicating cognizance of the work to be performed, the work location and conditions / restrictions, and approval to enter and perform the work.
- Copy 1 -(original) of approved RHWPs shall be posted at the entrance to the job location area to allow review immediately prior to entry to the worksite, Copy 2 shall be posted near the entrance to the RCA, and Copy 3 shall be retained by the RC&S0. Copy 4 of the RHWP <
shall be retained by the Work Supervisor.
. Exposure time sheets shall be provided for each RHWP, and all .
persons shall provide the appropriate information requested on the time sheet upon entry and exit from the work location.
. :All information entered on exposure time sheets shall be printed ,
legibly in ink. Information entered on RHWP time sheets indicates that the individual for whom information is entered has read, understands, and will comply with the requirements of the RHWP, and that entry and work- will be in accordance with established radiation protection rules and policy.
. Upon job completion, the individual (s) responsible fnr performing
. the-work shall inspect the work area to ensure that it is clear of materials, tools, equipment, or other items used or produced by l.', performance of the job. The individual (s) responsible for the work shall ensure that the work area is in a condition equal to, or better tFan, that at the commencement of the job, and the responsible supervisor shall sign the back of Copy 1 of the RHWP l posted at the jobsite, requesting termination of the permit.
l 2 11
+ The RHWP termination request shall be forwarded to the RC&SO and RE/HPS, who will verify the condition of the job site and sign approval-to terminate the RHWP.
All copies of the RHWP shall then be collected. The original (Copy 1) and Copy 2 shall be retained as file copies by the RE/HPS ' and RC&SO, respectively. The copies shall have the termination date ) recorded on them. After retaining a copy for the licensee,- all .. ! other copies should then be discarded. The licensee shall receive all required ANSI 15.10 documentation which will be specified in the S0W. 1 2.1'.4.3 Controlled Surface Contamination Area. Contaminated or potentially contaminated items, materials, and surfaces shall be handled, dismantled, and decontaminated within a Controlled Surface Contamination Area
-(CSCA). Radioactive waste material shall be placed in designated containers and stored in radioactive material storage zones. To minimize areas designated as CSCAs and the potential that contamination will be spread throughout these areas, smcll CSCAs shall be established to promote work efficiency.
These CSCAs may correspond to locations where cutting, dismantling, and decontamination operations are perforraed. When materials with loose surface contamination are properiy wrapped snd carefully handled to prevent breaking the wrapping, they may be corried through or handled in areas that are not - controlled for surface contamination. All work _ involving contaminated material shall be performed inside the boundantes of a_CSCA. - A Contamination Control Point (CCP), through which all . entries- and exits will be made, shall be located on the perimeter of each CSCA. The floor of the CCP shall be covered with paper, plastic sheet, or , other material. This is to provide an easily removab1e surface within the CCP to prevent the spread of contamination from the area. A step-off pad shall be placed at__the exit of the CCP. This shall be used when removing clothing during exit from the area. Receptacles for waste and contaminated clothing shall be maintained at the CCP. 2-12
Instruments for monitoring personne1' and equipment shall be on hand, All i
. equipaent, parts, materials, surfaces, and wastes that have been exposed to radioactive contamination or to neutrons from the reactor will be handled as radioactive anu shall not be released for unrestricted handling until they are surveyed and show results in compliance with NRC Regulatory Guide 1.86. If loose contamination is suspected to be in excess of regulatory limits on surfaces lnot accessible for measurement, the material shcIl be handled as 1 radioactive. Actual frisking shall be performed whenever possible in low-
- radiation background areas where audible response of the frisker ran be distinguished more easily. Adequately trained personnel will be permitted to frisk themselves.
Matarial tht! is neutron activated cr.d is to be retained shall be evaluated on a case-by-case basis as required by ANSI 15.10 and NRC guidance'. Radiation tags and labels will be available at the CCP to identify the contaminated or activated items being removed from the area. The entrance to , the contaminated arca shall be posted with:
- Approved RHWP specific to the operation ,
- Information concerning radiation and contamination levels
- Precautions for ertry
- Precautions for exit
- Step-off points F
- Frisking instructions i
- A. sign that prohibits eating, drinking, smoking, or chewing gum'or .
tobacco, 2-13 y v, . E-,e ,.4.. m. < , - . . - , . - , . , . , ~ . . . . , . _ . - - * - . . --
l The RE/HPS or a designee shall be responsible for the CCP and will ensure that personnel and equipment are surveyed ar.d logged as required. If work involving contaminated material inside the CSCA requires the use of glove boxes and enclosures, the installation, use, and dismantlement of the glove boxes will be supervised by the health physics personnel. , 2.1.5 Radiation Protection Facilities, Instrumentation, and Personal Protective Equipment Radiation protection f acilities, inst rumentation, and personal protective equipment used during dismantlement activities are discussed below. 2.1.5.1 Facilities. Facilities provided to enhance the effectiveness of the ORPPs will include the following:
. Facilities and equipment to clean, repair, and decontaminate personal prctective equipment, monitoring instruments, tools, and other material
. An emergency showet/ personnel decontamination facility
. Change areas to allow changing into anti-contamination clothing.
Anti-contamination clothing will be worn during work in the RCA, and the anti-contamination clothing will be disposed of as contaminated waste. There will be no contaminated clothing laundry.
. Control stations for entrance or exit of personnel into radiation or contaminated areas, for movement of radioactive waste material, and for movement of potentially contaminated equipment and instruments .'
. Equipment to facilitate communication between workers and .-
supervisory personnel between radiation and nonradiation areas 2-14
_ _ . . _ _ _ .- _. __ _ ._. _ __ _ _ _ _ - _ _. . ___._.m____. _.m . I l
+ Calibration source check facilities for the instruments that will be
-used during decommissioning.
l l Coordination shall be maintained with AMTL services, such as those provided by the local Fire Department, in accordance with the Memorandum of Agreement between_the AMTL and the town of Watertown, Massachusetts. Procedures shall be developed and included in an Emergency Response Plan pnor to any decommissioning activities. 2.1.5.2 Instrumentation. A wide range of portable and nonportable l instruments and lab-cuunting equipment will be supplied by the decommissioning contractor and used during decomissioning for radiation surveys, radioactive contamination. surveys, personnel monitoring, area monitoring, air monitoring, and sample analysis. All- instruments shall be calibrated in accordance with the specifications ! i contained in ANSI N323-1977 or the most recent revision, Detailed calibration l records-(including date, method, source description, results, and person) i shall be kept as quality assurance records and will be auditable under a quality assurance program.
-On a daily basis, or as frequently as required, each type of
~
instrumentation shall be-checked and source checked to verify that it 1s functioning properly and is in calibration. Table 2-3 lists typical types of instruments required for a
-decommissioning project.
2.1.5.3 Personal Protective Equipment. Other personal protec'tive equipment shall be provided by the contractor for use as needed. Typically, such equipment includes:
. Anti-contaminaticr. clothing I
2-15
r
- Contamination control equipment, such as hoods, plastic containers, bags, filters
* . Signs, labels, tags
+ Special. tools *
. Decontamination equipment
- Mobile or temporary shields
- Respiratory protection devices-
- Hard hats, steel-toed boots, and glcves.
Donning and removal of anti-contamination clothing shall be limited to designated change areas and RCA exit area.-
-Respiratory Protection Program: A respiratory protection program in compliance with 10 CFR 20.103, ANSI Z-88,2, NRC Regulatory Guide 8,15, and OSHA shall be devaloped by the decommissioning contractor to provide protection against airborne radioactive and/or chemically-hazardous substances. The following elements are included in the program:
- Written standard operating pra edures governing selection and use of respirators
+ -Assignment of responsibilities a Types of records .-
- Training of employees and supervisort , .
+ Quantitative and qualitative testing
=
Work area surveillance l
- Medical surveillance 2-16 i
Table 2-3. Typical radiation survey and monitoring instrumentation and equipment to be provided by the Decommissioning Contractor
. Fortable ion chamber rate meters
,,.
- Portable GM survey meters
. Alpha survey meters
. Pocket ion chamber dosimeters
. ' Area monitors including periphery area monitors a -Air sampling equipment a Windowless gas flow GM counting systems
. -Liquid scintillation counter system
. Hand and foot monitor
. Pressurized ion chamber a Gamma spectrum analyzer
. . Permanent personnel dosimeters, either film badges or thermoluminescent dosimeters"
.a. All radiation workers, in accordance with AR 40-14, shall wear Army radiation dosimetry badges at all times during decommissioning.
Thisagill be in addition to the contractor required badges. Al? dosimetry shall be NVLAP accredited. 4 2-17
+ Special respirator use problems and limitations
- Maintenance and repair of respirators.
The contractor shall supply only respiratory protection equipment , approved by the Mine Safety and Health Administration and the National Institute for Occupational Safety and Health. Potential exposure hazards, radi.oactive particulates and vapors, and airborne toxic chemicals shall be monitored and evaluated. Workers shall be instructed and supervised to ensure N that all respiratory protection equipment provided is used in accordance with the training and instructions received. The contractor shall routinely inspect _ respiratory equipment, and protect it from outside contamination and damage. The RC1SG dali delegate responsibility to the Certified Industrial llygienist (C'u) to direct, evaluate, and provide guidance on all aspects of the facility respiratory protection program and ensure that employees required to wear respiratory protection are physically and medically able to do so. The Respiratory Protection Program shall ensure that exposure of individuals to concentrations'or radioactive materials in the air is in compliance with 10 CFR 20.103. Tht respiratory protection program administration shall be accomplished under the guidance and direction of the ClH in coordination with the RC&SO. The AMIL Technical Monitor will provide oversight for the respiratory p.otectioT program. In addition, the CIH shall be responsible for maintaining an adequate _ supply of respirators and cartridges on-site for personnel use.
-All purchasing of respiratory equipment will be under the guidance of the CIH.
Respiratory selection shall be based on the following criteria: J
- Nature of the hazard .
-
- Physical properties of the contaminants involved
+ Contamination on surfaces and airborne contamination 2-18 'I-
.: ) 4
+ Location of the hazard J
- Time frame-for which respiratory protection will be required
- Operational activities of personnel required to wear respiratory.
equipment
- Functional capabilities and limitations of respiratory equipment
- Potential for the presence of condition:; immediately dangerous to life and health
+ Potential for the presence of oxygen deficient atmospheres.
All decommissioning personnel required to use respiratory protection shall receive period _ic training pertaining to all respiratory protection. This training shall be given under the guidance of the CIH, Training shall include but not be limited to the following: '
- Proper use of all available respiratory protection including e hands-on training
+ Reasons for the selection of a particular type of respiratory equipment based on potential hazards I
i Functional capabilities and limitations of all available respiratory. ~ equipment f
;. . . -Identification .of respirator malfunction and how to correct the malfunction t
~,
Proper methods'of donning respiratory protection equipment
- Reasons for determining respiratory fit, methods to be used, and factors affecting respiratory fit 2-19
- Proper care and maintenance of respiratory protection
- Training in the use of respiratory protection as it relates to the recocnition and handling of emergency situations
. Discussions of potential contaminants against which the wearer is to .
be protected, including physical properties, physiological action, toxicity, and means of detection ..
- Discussion of the application of various cartridges and canisters available fer air respiration
. Instruction in emergency action to be taken in the event of malfunction of the respiratory protection devices, All personnel required to use respiratory protection shall be advised that they mdy leave the work area at any time for relief from physical or psychological distress, procedural or communication failure, significant deterioration of operational conditions, or any other condition that might require such relief.
The CIH shall designate which personnel are qualified to give instructions on respiratory protection. Qualifications. shall be based on the designee's knowledge of the application and use of respiratory protection and the hazards assor,iated with potential chemical and radioactive contaminants. All decommissioning personnel required to wear respiratory protection and who have demonstrated that they are physically and medically able to do so shall receive a qualitative respirator fit test. to be administered under the . guidance and direction of the CIH. The program shall include employee orientation, employee sensitivity tests, performance of the fit test in a test enclosure, and respirator assignment. The qualitative fit test protocol will use Isoamyl Acetate or Irritant Smoke tubes as appropriate. Records of respirator fit test results shall be provided to the licensee.
?-20
l i
.J
)
In addition, all: personnel shall be instructed in the proper procedure - i for the performance ~ of the posi_tive and negative pressure tests, These quick .j respirator -fit checks shall be performsd by all personnel immediately after _ donning' approved respirators and prior to entering an area designated for l respirator use. l
.- q Personnel shall only wear air purifying respiratory equipment for which they have successfully passed the qualitative fit test protocol, No respirators shall be worn'by personnel who have facial hair such as beards or long sideburns which interfere' with the sealing periphery of the respirator face piece or with respirator valve function, j i
L Contact lenses may be worn by any employee while in an area designated for respirator use. Prescription glasses may be worn as long as the seal of l :the respiratory face piece to face is not directly affected,- .) e L Only the CIH or a properly trained designee shall be permitted to issue l 1 l o respiratory protection to decommissioning personnel or outside contra-tors and i j . visitors. I All subcontractor personnel and visitors shall be required to adhere to L the same respiratory protection procedures as regular decommissioning personnel.
^
For unusual operational instances or special projects, respirator i issuance shall be made under the guidance and direction of the CIH. Al'1 workers for whom the potential of contact with hazardous materials
". ; exists.shall participate in a medical surveillance program. _As a minimum, this program must provide baseline health assessments to investigate existing conditions that may predispose a worker to illness following exposure to .
hazardous substances or to the physical demands of using protective equipment, t In addition, periodic health assessments shall be provided to screen workers for signs of occupational exposure to toxic agents and to determine their subsequent assignments. 2--21
2.2 INDUSTRIAL $AFETY AND HYGIENE PROGRAM The Industrial Safety and Hygiene Program (ISilP) for the decommissioning project is concerned with the protection of all personnel from pot;ntial nonradioactive exposures and hazards. The ISHP shall be developed by the , decommissioning contractor, and the ISHP shall include a hazard analysis, in addition, the ISHP shall specify the applicable regulations, standards, and other requirements to be followed including the U.S. Army Corps of Engineers Safety and Health Requirements Manual, EM 385-1-1. The ISHP shall be administered in accordance with OSHA regulations, in the absence of a particular regulation, gaidance shall be obtained from the NIOSH or the American Conference of Governmental Industrial Hygienists (ACGlH). 2.2.1 Personnel The ISHP shall be administered by the contractor's CIH in coordination with the RC&SO. In addition to responsibilities previously discussed as aspects of the RPP, responsibilities shall include:
. Inspectio,. and audits
= Occupational health Medical surveillance Hearing conservation First aid
- Emergency services
. Operational activities.
2-22 .o
i 2.2,2 Training To su;plement the comprehensive training program described in Section 2.1.2, the contractor shall also supply all workers with instruction concerning the project safety program through orientation / training prior to being-assigned to project activities. Each new hire or transferee shall f attend one of these orientations,'which consists of instruction in job safety i action plans, hazard recognition and correction, fire extinguisher training, and safety awareness films. Specialized training applicable to specific conditions shall_be given as the progress of decommissioning activities , mandates. Supervisory safety training is an integral part of the safety training program. Supervisors shall receive a-safety orientation detailing the safety responsibilities of their positions. L Training courses and a qualified staff roster shall be documented and updated, with-follow-up training conducted as needed. Topics for presentation 4 at these ISHP training sessions shall include:
- Specific project safety procedures
. Fire protection and prevention
* - Work practice. procedures ard tool-box safety
- Confined-space entry
. Special housekeeping requirements
+ -Material-handling; techniques
- Safety'and warning devices I
Hazard-identification and reduction
-Hazard communication 2-23 i
J-_-.___________.______i____ ____.2_ _ _ _ _ _ _ . . _ _ . _ -
,-.--_,,,w,,.---,x. -_ .,-e-, , , . - - - . , . , . . ~ , _,, ,y ... _ ,,
- Enforcement _ policy e
first aid and emergency procedvres and equipment, Records shall be kept of all personnel attending, level of ., accomplishment, follow-up se';sions, etc., at necessary, to ensere,that the sppropriate awareness and competency have been demonstrated. ,, t Individuals performing asbestos sampling shall have completed the Asbestos Hazardous Emergency Response Act (AHERA) training course for asbestos inspectors 2.2.3 Administrative and Work Practice Controls These controls comprise the measures taken to limit chemical exposure and safety hazards and to reduce the risks associated with the decommissioning activities. Essential to the ISHP are the RHWPs which are discussed in Section 2.1.3.2 and tecess controls discussed in Section 2.1.3.3. 2.2.3.1 Exposure Limits. Personnel exposures to toxic / hazardous materials shall not exceed limits established by OSHA or those recommended by ACGlH in " Threshold limit Values and Biologica! Exposure Indices " -
, Evaluation of potential toxic / hazardous risks to workers shall be
_ performed 'oy the CIH using appropriate methods. These methods may include the use of_ Dragger tubes, air sampling, analyses for hazardous materials, and visual observation based on experience and training. In the event that the evaluation by the ClH reveals potential
- toxic / hazardous risks to workers, the CIH will require appropriate personalz protective equipment.
2.2.3.2 Inspection and Audit Programs. _ Inspections shall be conducted
-routinely by the contractor's ClH and RC&SO during active work phases and on at least a weekly basis. Safety violations shall be recorded, identified, and 2-24 L__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . - _ - _ - - _ - _ -_-._ .-
- _ _ . _ . . - .m.____.-.. _ -m.-._ __._ _.._ _ __ _ _._ _ _ _ . _ _ _ _ .-_. _ _ _
corrective actions taken ihvediately. Copies of any infraction _ notices shall be maintained by the RC&SO ai.d documented in a Weekly Safety Report to the Army's KOR. The following items shall be specifically addressed during these routine inspections:
- Barricades
. Safety signs
. Scaffolds
. Hoisting and rigging
. Confined-space entry
. Excavations
. Torch cutting
. Hearing protection
. Any'other industrial hazard Radiation / Hazardous Work Permits-Personal protective equipment _(PPE).
Lin_ addition-to inspections conducted by the contractor, the Army wili
. _ provide an independent Quality Assurance-Evaluator / Health Phy:icist (QAE/HP)-
to continuously monitor-the project to ensure that decommissioning is performedLin compliance with all specifications and requirements. The independent QAE/flP will report- directly to the licensee. Supervisors s'nall be required to participate actively in the investigation of any accident occurring in their areas and which result in any. 2-25
~ ' '
-_--E,,-- ..-.E,<~.-- E. ,'-_E - - - + . ,, --,e + ~ - " ww
personal; injury to employees under their direction, equipment or property damage, and near misses with the potential for serious injury or loss. The investigation will'be aimed at determining facts, not fault, so.that recurrences =can be prevented. , e In order to provide verification of the program, an audit procedure of ,
- the ISHP will be Jeveloped by_ the licensee, incorporating approved evaluation criteria. Audits will be conducted by the independent QAE/HP. Audits are conducted for:
- Compliance with all safety requirements
. Implementation of haalth and safety procedures
- Health and safety organization
+ Job descriptions and tasks
+ Review of records and docunientation relating to health and safety
+ SiteLlayout and inventory
- Training materials.
The audit' criteria include, at a minimum, the evaluation of:
+ Written procedures
+ Qualifications = education, and training of management and staff
- Cois;unications and coordination of the various health, safety, and .
medic?? 6?emants of the. program
+ - Environmental surveillance
- y-2s W
, - , ,<,w - -,e - -
,,-r- ---- ,-+,,e = y- --w ,
M. 4 p e
- Facilities, apparatus,_and monitoring equipment i
i
- Medical surveillance i l
- Emergency planning.
- Using this procedure, an assessment will be completed and the results j properly communicated to the RL&SO and the licensee.
j- Written recommendations shall be prepared to improve deficient areas. It l is the responsibility of the contractor's Project Manager to ensure correction of deficiencies and documentation of actions taken. 1 i i 1 2.2.3.3 Accident Reporting. Accidents resulting in a fatality, lost- { time injury or illness, hospitalization of 5 or more personnel, or property j damage to government or contractor property (which occurred during the ] performance of the contract) equal to or exceeding $2,000.00 shall be ! _ telephonically reported to the NED Safety Office and COR, as soon as possible, l but'not later than 2 hours after occurrence and reported in writing _within 5 4 days of occurrence.on ENG Form 3394 (encl)3. All other accidents / incidents
- shall be telephonically reported within 8 hours of occurrence. All accidents l that occur during decommissioning of the AMTL Reactor shall be reported in compliance with EM.385-1-1. The COR will provide all information regarding i accidents to the.MTL Safety Office.
l c 2.2.3.4 Medical Surveillance Program. A medical surveillance program I shall. be established by the contractor for all workers who may be
, occupationally exposed to radiological or hazardous chemical agents. The 5
- i. program shall comply with applicable state and Federal requirements including
. 000 6055.8, Occupation / Radiation Protection Program and may include'the following items as appropriate in the baseline health assessment:
. Occupational history i
7- 2-27 1 1
. Medical history
. Family history
. Physical examination ,
Pulmonary function testing g .
. Audiometric testing
. Baseline bioassay.
The regulatory 9t X9 for the baseline bioassay is NUREG 8.9. ANSI standards shall be used to determine maximum internal doses. Criteria for the health assessment shall be developed to identify the need for pre-employment and periodic health assessments, termination examinations, and return-to-work and other special examinations. The decornmissioning contractor shall determine available clinics and hospitals for routine and emergency medical needs. Medical records shall be maintained, with af tention given to federal requirements, inclusion of exposure data, appropriate update frequency, and , access privileges. The work site shall be monitored for health hazards associated with the work environment, including chemicals that may be present in liquid, dust, fume, mist, vapor, or gaseous forms. Physical ha'ards such as noise, ' pressure, vibration, and illumination will also be monitored and controlled. 2.2.3.5 Hearing Conservation Program. A hearing conservation program shall be established by the contractor for all workers who are exposed to noise levels in compliance with U.S. Army Corps of Engineers Safety and Health Requirements Manual, EM 385-1-1. This program shall include: 2-28
- _ . _ m _ _ . . _ _ _ _ _ - _ _ . _ _ . . . _ . _ . _ _ _ _ _ . _ _ _ - . _ _ . _ _ ,
=
t
.. . Noise monitoring in er eas where the levels exceed 80 dBA I
Audiometric testing- fo: # workers (as part of the medical surveillance program)'to determine baseline hearing performance before exposuie and test results after exposure
- Personnel training and educatior.
.- Recordkeeping a Hearing protection devices.
Personnel who are assigned tasks.in known noise-nazardous areas as defined in EM 385-1-1 shall be enrolled in the hearing conservation program 1 priot to beginning their work. Noise' control measures in compliance with EM 385-1-1 shall be determined by the ClH after appropriate noise monitoring is completed in the work area. The contractor shall maintain records that documei t all noise monitoring conducted, employee training done, control measures implemented, and.
- protective equipment issued.
2.2.326 Fitness for Duty Program. The decommissioning contractor shall implement an apprenriate Fitness for Duty Program to comply with NRC Notice Number 90-81. The requirement and procedures to implementLa Fitness.for Duty L Program will be incorporated-into the Statement of Work. The program will l . assistithe AMIL Commander in providing a drug-free work-place. 2.2.4 Operational Activities l -i The requirements for fire protection, emergency response, and equipment / tools areLdiscussed below. 2-29
2.2.4.1 Fire Protection and Prevention. Fire protection devices shall be made available by the contractor in appropriate locations during decommissioning tasks. Portable Type A and B/C fire extinguishers shall be strategically located to serve areas partitioned by the various decommissioning activities. Fire prevention measures shall be implemented to a"oid ignition hazards from electrical wiring and equipment and from combustible materials. Smoking - shall not be permitted in areas where a potential fire hazard is present. 2.2.4.2 Emergency Response. Prior to start of decommissioning activities, the decommissioning contractor shall prepare an Emergency Response Plan in compliance with EM 385-1-1. The Emergency Response Plan will be approved by the licensee and COR. 2.2.4.3 Hand and Power Tools and Cutting Equipment. The condition of the hand and power tools used during decommissioning shall be rs tineiy checked for proper operation and for compliance with applicable requirements. 2.2.4.4 Lifting Equipment. Lifting equipment (such as hoist, fork lift, and bridge crane) used in decommissioning shall comply with applicable provisions of OSHA. These provisions include:
- Compliance w.th the manufacturer's specifications and limitations applicable to equipment operation
- Posting of rated load capacities, operating speeds, and special hazard warnings or instructions
- Inspection of equipment by competent personnel prior to each use and during use to make sure it is in safe operating condition .
- Limiting the travel of rail-mounted equipment with limit stops Removing from service any equipment which has damaged wire ropes, chains, or other c.ompcnents.
2-30
j 2.2.5 Personal Protective Measures To minimize the effects of industrial and radiological hazards associated with decommissioning, specific health and safety measures shall be implemented by the decommissioning contractor. Anticipated hazards and mitigation measures are listed in Table 2-4. The decommissioning contractor shall
- specify task-specific personal protective equipment in the work procedures for each task and in the RHWP covering each task.
2.2.6 Excavations Excavations rereired during decommissioning activities shall comply with applicable provisions of OSHA and EM 385-1-1. These provisions include:
- Sides of excavations tapered in compliance with OSHA requirements
- Protection of workers with Personal protective devices as discussed in Section 2.2 of this document.
+ Provisions to prevent workers from standing under loads handled by I
lifting equipment
. -Daily inspection of excavations by contractor industrial safety personnel for evidence of potential or actual cave-ins or slides
.
- Supporting systems (e.g., underpinning, etc.) designed by qualified contractor personnel-and inspected daily p '. - Excavated materials ano other material stored at least 2 ft from the edge of the excavation -
- When using heavy equipment in the vicinity of excavation, the sides
- of the excavation braced to resist extra pressure by superimposed loads 2-31
Table 2-4. Mitigation and monitoring of hazards during decommissioning Anticipated Hd1EdJ' liiU9alhlt HertMpdfit ___ i borae
- Respirators (full face) + Whole-body counts HEPA filtration units Continucus air sattpling r;dionuclides + * -
+ Grab air sampling falling and + safety glasses with + Incident Reports >
flying debris side shields
- Limited access, safety shoes
- Hard hats High sound + Ear protectors + Physical exams levels e dB measurements Betegamma e Limited access
- personnel dosimetry exposure rates
- portable shielding + itally surveys High heat and
- Ventilati]n + h ,nperature measurements humidity + Work breaks + Stay time limits
+ Cooled protective clothing loose surface
- Anti-contamination + frisking contarnination clothing
- Daily surveys
- Citanup/ decontamination Airborne hazardous
- Respirators and supplied
- Air sampling chen.icals/ vapors air
- Cleanup / decontamination Airborne dust
- Water fogger + Air sampling
- Safety goggles
- a. The listed anticipated hazards may be encountered during all decommissioning tasks discussed in Chapter 3.
4 2-32
i
+ Adequate barrier physical protection orovided around the excavated area ;
- Safety harnesses / belts provided for use in confined spaces Ladders and scaffolding used in excavation provided around the i excavated area
+ Ladders and scaffolding used in excavation complying with the ,
applicable provisions of OSliA. l 2.3 CONTRACTOR ASSISTANCE A decommissioning contractor will be used to perform the decommissioning of the AMIL Reactor. The activities to be performed by the contractor include the following:
- Performing all decommissioning operations
- Supervising day-to day decommissioning activities including directing .
craft supervisors and crew leaders .
. Providing health physics support including install'ng, calibrating, and testing equipment; and conducting radiological surseys
- Providing health and safety support, including a Clfi
+ furnishing quality assurance support, including preparing procedures ,
and complying with quality during decommissioning
- Providing crafts and labor for temporary construction work, performing decontarination and demolition tasks, and processing, packaging, and shipping rac'ioactive mater'als.
t 2-33
___m_- e I i
! i These will be ongoing activities during the entire decommissioning period, !
4 with the Army's personnel oserseeing and reviewing the work as it takes place. However, the contractor will be delegated responsibility for health and safety via the contract between the Army and the contractor. , i f i [ t
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i 1 l CHAPTER 3 DECOMMISSIONING TASKS, SCHEDULE, COST, AND FUNDING
-l
- 3. INTRODUCTION 1
l Major tasks to be performed during decommissioning of the AMIL reactor facility are shown in figure 3-1, Work Breakdown Structure (WBS) for l Decommissioning the AMIL reactor. Each element in the WBS will be discussed in Sections 3.1 through 3.6. The estimated schedule and cost for the project are given in Section 3.7. The source of funding ior the project is discussed in Section 3.8. Descriptions of decontamination tasks include general procedures for how the facil'ty should be decantaminated; the descriptions address health and ; safety coit lderations if applicable. Task descriptions in this chapter are general, but detailed work instructions will be specified in work procedures to be prepared by the decommissioning contractor and be available for review ' by the AMTL staff. I All operations and tasks that involve equipment and materials that produce ionizing radiation shall be conducted in such a manner as to maintain radiation exposure to personnel as low as reasonably achievable (ALARA). Operations and tasks involving ionizing radiation shall be planned so that the limits established by the NRC, DA, AMC, OSHA, and MIL regulations are not exceedec. 4 W;rk procedures will also specify radiological surveying procedures to be followed during removal and disposal of equipment, components,. material, and j other items. These surveying procedures are required to ensure that alI l *.- radioactively contaminated equipment, components,. material, and other items are disposed of appropriately, it is essential that no radioactive waste ) above releasable levels, as defined in Reference 4, be disposed of as domestic waste.
- 3-1 i
i
I f i 3.0 i 4 AMTL Reactor [ Decommissierdrg I I ! t r 3.1 32 J3 3.4 3.5 36 [ Site Preparation Probet .. Aux Structures ' Reactor Building Termiraton BackI21 j Management Removal Decontamatation Survey and Grade 3.1.1 3 2.1 3 3.1 3.4 1 3 5.1 ] Buildsq 100 _ Statement M Cistern 242 PoolIntemals - P*:aseI of Work and Anrufus 3.1.2 3.3.2 3 5.2 322 3.4.2 Building 97 _ Building 97 - Phase 11 Decommissioning Poing d Pool Liner l f 313 Contract 3 4.3 ' Auxiliary 3 3.3 PWom , Structures 3 2.3 Secondary Onsite Monitoring Coolant System 3 4.4 i y 3.1 4 _ Basement d MisceHaenous l- 3 2.4 Electrical 3.1.5 ress Ws 345 , Setup Trailers & 3 2.5 - t- Staning Areas I & Equipment 1 , Final Report i 346 ) 3.1.6 i ,- 3.2.6 - Basemer:t Sumps Establish j ridacts 347 Barriers - Stored / Displayed Gamma Faci!cy
- & Storage Tubes j 348 Pool
- 349 t
)- _' Buildag intemal 1 De::entamination
- j Tu esos 1
Figure 3-1. Work breakdown structure (WBS) for deconunissioning the AMTL ] 1 Reactor i l
i Surveying will include smearing for removable contamination, and direct 4 contact radiation measurements for alpha and beta-gamma radiation. for direct radiation measurements, the scanning speed will be slow enough to ensure detection of the most restrictive release levels specified in Table 1 of l Reference 4. The scanning speed will depend upon specifications of the instrument being used. Tb scanning speeds will be calculated and included in i
.. the procedures.
Prior to decommisst ming, ;'a "e.s listed below shall be performed: lasks to be Prrformed Prior _w.ggfsmissioning i
- 1. Ensure that all experiment equipment and materials are removed from
.the recctor building, including all concrete and phenolic blocks, cabinets, desks, etc.
- 2. See that all concrete blocks and other materials stored around the secondary coolant pumps and sump are removed prior to contractor :
mobilization.
- 3. Enscre that Cistern 242 and the secondary coolant sump are empty and that no controllable inflow of water exists. ,
- 4. Provide utility hookups for the contractor. These must include water, power, telephone, and sanitary sewer lines.
- 5. Provide office space for the Quality Assurance Evaluator (QAE) in th'e immediate vicinity of the reactor building. This office will include a desk, chair, file cabinets, lighting, power, and telephone service. ,
~
- 6. Designate sufficient area for the contractor to place his equipment and portable facilities, and to stage materials awaiting use or disposal.
1 3-3
.m.,,._-.___.m..a._,.._.._ . _ _ _
- 7. prepare the reactor building and immediate area for decommissioning operations by performing the actions listed below,
- a. The areas in Building 97 adjacent to the airlock entry into Building 100 and the pipe alley containing reactor related piping .,
will be free of equipment and materials to allow the contractor free access.
- h. Remove two cabinets adjacent to Building 100 in the area of the secondary coolant pumps,
- c. Remove the yellow steel structure adjacent to Cistern 242.
3.1 SITt PREPARATION This section briefly describes the activities and tasks required to prepare the AMIL reactor facility and site for decommissioning. Tasks described in Sections 3.1.1 through 3.1.4 are not part of the decommissioning contractor!s scope of work and will be performed by the licensee prior to start of decommissioning. Site-preparation tasks described in Sections 3.1.5 .and 3.1.6 shall be performed by the decommissioning contractor. 3.1.1 Building 100 All experimental equipment, office furniture, file cabinets, concrete blocks, nonradiologically contaminated lead bricks, and all other nonreactor . related materials will be removed from the building and the surrounding area, t Reactor control equipment that is reusable will be sent to other reactor , facilities prior to initiation of decommissioning operations. , 3-4
3.1.2 Building 97 The brass plaque above the airlock door will be removed and placed in the artifact repository. 1he areas adjacent to the airlock and piping corridor will be cleared of
. . all loose mater ai .
3.1.3 Auxiliary Structures Cis+ern 241 and the secondary coolant sump will be emptied and isolated f rom all sources of liquid influx other than groundwater. 1he sources of any fluids that are controllable will be pernanently isolated as close to the source system as possible and identified. A yellow metal structure adjacent to Cistern 242, the two metal cabinets and pallets of concrete bricks near the secondary coolant pumps, and any other loose items in the area of the reactor will be removed from the vicinity. 3.1.4 Hiscellaneous 4 Areas within AMTL and adjacent to the reactor building will be designated for the contractor to place temporary facilities and to stage materials and wastes. A buf fer zone will be established by erecting a fence around the reactor building to prevent decommissioning and other base activities from impacting each other. The approximate location of the fenced area is shown in figure 1-2, A protected storage area will be designated by the licensee and prepared for the temporary storage of containerized low level radioactive waste until the waste can be shipped. A temporary office for the independent QAE/HP will be established as close to the reactor building as possible. 3-5
Provisions for temporary utilities such as telephone, power, water, and sewer for use during decommissioning activities will be identified. 3.1.5 Setup Trailers and Staging Areas rollowing the award of the cont act for the conduct of decommissioning, the decommissioning contractor will mobilize the equipment and facilities .. needed to support the decommissioning activities and locate them at AMIL. It is anticipated that the contractor will bring office trailers, change room / shower facilities, decontamination facilities, radiological analysis j laboratories, various pieces of construction / demolition equipment, barricades, ' signs, and other items to support their activities. The contractor shall also establish areas within the fonced area for the i staging of materials prior to utilizd lon. Packaged waste shall be l temporarily stored (if necessary) in i protected storage area near the reactor i facility prior to shipment to a disposal site. In each case, the amount of materials staged or stored will be kept at the lowest level possible. The decontamination and change facilities shall be located as close as i practicable to-the containment building to minimize the potential for the i spread of contamination during access and agress. At the completion of the contracted work and following acceptance by the licensee, the contractor shall ensure that all materials and equipment , belonging to the contractor are removed from the AMIL. i 4
-3.1.6 Establish Barriers *
._In order to maintain separation of decommissioning activities from the-b t
other activities at AMTL, it will-be necessary for the contractor to establish
- barriers. The barriers shall be established to include the minimum area
-needed by the contractor for isolation of the decommissioning' activities.
Barrier locations will be specified in the 50W. To minimize impacts of restricting access to areas of AMTL it may be necessary to relocate the 3-6 i f
l l i barriers to acconnodate the various stages of decommissioning. Any relocation must be approved by the AMIL Connander. At the completion of portions of the work scope requiring barriers, the barriers shall be removed from AMTL promptly. 3.2 PROJECT MANAGEMENT i j The decommissioning operations will be performed by a decommissioning contractor. The overall project management, however . will be accomplished by the U.S. Army Corps of Engineers (COE), New England Division (NED). This section describes the tasks required of the COE, NED. i 3.2.1 Statement of Work A Statement of Work (S0W) will be prepared for and approved by the Army. The 50W will be the basis for which the decommissioning contract is written and selection of the decommissioning contractor is made. The 50W will specify the work to be performed by the decommissioning contractor, standards of performance, and all deliverables to be supplied by the contractor. The 50W must also include a detailed description of the decommissioning Contractor's responsibilities for health, safety, and environmental implications and effects of the work performed. The 50W will also state the items to be furnished by the Army during decommissioning. The required work specified in the 50W will include all decommissioning tasks described in this DP, except those tasks specified to be performed by the Army i or an independent contractor. 3.2.2- Decommissioning Contract Following completion of the 50W, a Request for Proposal (RFP) will be prepared, proposals requested and evaluated, and a contract awarded for the L performance of decommissioning as described and specified in the 50W. The RFP 3-7 i
1 will include the 50W, which will be the basis for the proposals submitted by prospective contractors. The contract will be for the work described in the S0W.. Prior experience with decommissioning of an NRC-licensed facility, as well , i as health physics and safety experience, shall be a requirement for any l contrattor chosen. A complete list of qualifications for the decommissioning i contractor will be included in the 50W. i 3.2.3 Onsite Monitoring i The Army COE will provide a QAE/HP to be onsite at all times while the contractor is performing work. The QAE/HP will ensure that the contractor is performing work in-accordance with the terms of the contract. In addition, the QAE/HP will be immediately available for technical consultation when unforeseen situations are encountered. The QAE/HP must have decommissioning experience. The QAE/HP will provide written daily status reports to the designated Army point of contact as to the status of work progress as compared to the plan, problems, corrective actions, and any other pertinent information. Should any high-concern incidents occur, such as injuries, contamination Lreleases, etc., the QAE/HP will immediately inform the Army contact of the situation. The QAE/HP will also maintain a photographic record of the work progress fer inclusion in the status reports and final report. L3.2.4 Progress Reports ? Written daily _ progress reports will bo 6 de by the QAL/HP to give the ' status of the work with respect to the plan, report on any unusual occurrences, and status of any corrective actions. These reports will be L -brief and concise descriptions of the work performed each day. 38 l _. ~ . _ u _ & .. _._..__.._.a...._____._..- .__ __ _ . _ _ _ _ _
i Monthly reports will also be prepared by the QAE/flP. lhese reports will be detailed status reports of the technical, budgetary, and schedule progress of the activit ies. This report will assess how work is going as conpared to the contract schedule and will provide variance analyses, mpacts of variances, and status of corrective actions. Copies of all progress reports will be distributed to the LOI., ! , (OR, and the AMIL Commander, as a minimum. 3.2.5 Final Report A final Report will be prepared by the QAE/ lip to document decommissioning of the AMil Reactor. This report will include descriptions of the tasks completed, the actual schedule of accomplishments, photographs of the work and final conditions, final radiological ter.ditions, a summary of worker exposure, waste volumes generated and repositories of the wastes, and lessons learned. 3.2.6 Artifacts Stored / Displayed Artifacts from the reactor facility that meet radiation release criteria specified in Reference 4 will be collected by the Army prior to start of decommissioning. The Aizy may choose i; di .1,y a 0. - , , . the artifacts to help describe activities that were condu 'ed at AMil throughout its history. Examples of artifacts that will be preserved are the model of the facility, the plaque over the airlock in Building 97, photographs of construction and operations, and operating logbooks. 3.3 AUXILIARY STRUCTURES REMOVAL 3.3.1 Cistern 242 This below-grade concrete cistern (Figure 3-2) is to be removed by demolition and any contaminated soils surrounding the cistern also removed. 3-9
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The removal of the cistern is to be one of the first activities performed under the decommissioning order so that information concerning the immediate subsurface conditions at the cistern may be applied to any future excavation and removal activities, prior to demolition, the cistern is to be sampled f or radioactive and hazardous materials. in the unlikely event that radiological contcmination is found inside the cistern, the inside of the cistern will be decontaminated to releasable levels before the cistern is demolished. Iollowing receipt of sample analyses data, the cistern will be demolished and the materials disposed of in a manner approved by the AMil as determined by the analyses results, it is anticipated that the cistern will contain little or no contamination above releasable levels, lhe cistern demolition will include removal of the chainlink fencing, pump, and electrical supply equipment, followir.g removal of the cistern structure, the soils surrounding the cistern will be radiologically surveyed, and any contaminated soils shall be immediately removed and packaged to prevent contamination spread, if immediate removal of contaminated soil is not feasible, the soil will be covered to prevent contamination spread. When the contractor is reasonably ensured that the soil is at releasable levels, phase 1 of the final survey will be performed as specified in Chapter 8.
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Removing the Cistern will also include removal of piping connecting the cistern to Building 97. The piping removed as part of this task will be radiologically surveyed, results compared with Reference 4 and disposed of as clean waste or radiological waste. During entry into and while conducting work in enclosed spaces, special
, precautions specified in Section 2.2 will be implemented to ensure the safe conduct of the work. A permit for confined space entry must be obtained and continuous monitcring conducted for oxygen-deficient / explosive atmospheres.
Adequate sicping and/or shoring of the sides of the excavation will be implemented to minimize the potential for cave-ins. Due to the nonuiiform nature of the fill materials in the area of the reactor building, sf oring will be the preferred method of cave-in prevention. 3 3-11
3.3.2 Builning 97 Piping There are several piping runs remaining in lluilding 97. 1hese are liquid waste lines leading to the previously removed waste plant, demineralized water lines, steam and condensate lines, and water supply lines to the cistern. , prior to removal, these lines will t'e checked to ensure that the systems are not pressurized, the lines drained of all materials, and verified to be free " of asbestos containing insulation materials. Should asbestos-containing materials be encountered, the decommissioning contractor will follow appropriate procedur(s prepared b/ the decommissioning contractor and spproved by the Army. Asbestos sampling operations shall be performed in accordance with U.S. Army Technical Manual 5 612. If asbestos work is performed, only qualified asbestot workers will perform the work. When the piping has been determined to be empty and free of external contamination, the piping will be removed, sectioned, and the interior surveyed to determine if the individual sections can be disposed of in accordance with the unrestricted use criteria specified in Reference 4. 3.3.3 Secondary Coolant System The secondary coolant pumps remaining atop the sump are shown in figure 3-3. They will be removed. surveyed for radioactive materials, labelled, and staged for disposal. The sump will be surveyed, excavated, and removed in the same manner as riescribed for Cistern 242. During the removal of the secondary coolant sump, underground piping and electrical conduit between this region and the reactor building will be excavated, surveyed, and appropriately disposed of. lhe sump contents, sump, and underground piping are not expected to be contaminated. 3-12
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l 3.4 REACTOR Bult. DING DECONTAMINATION During reactor building decontamination, components will be removed and contaminated sections separated from uncontaminated sections. In no case will the facility cortainment be broken until it has been demonstrated that the , facility has been decontaminated to releasable levels. 3.4.1 Pool Internals and Annulus To minimize the radiation fields to be experienced by decoh,mtss ionirg workers in the containment building, the first step to tie taken should be the removal of loose materials in the annulus, the core support structure, and beam tube ends in the reactor pool area (in that order). Because of the high radiation fields associated with the annulus materials and core support structure, the contractor shall use remote cutting and handling equipment during this removal to minimi7e worker radiation exposure. Inniediately af ter renioval, the items with high radiation fields shall be appropriately shielded during packaging for disposal. The deluge system piping will be removed f rom the reactor pool area to provide access to the equipment within the pool area. The beam tube ends will be removeo by cutting of f the portions that protrude into the pool. These pieces will be placed in waste containers and staged for disposal. Instrumentation, remote operating equipment, access ledder, and manway cover to the lower pool access opening will also be removed as part of this task. The last components to be removed from the pool will bc the doors that - connect the annulus to the reactor pool. These will also be placed in waste containers. 3 14
_ . . _ _ _ _ _ . - ~ _ . . _ _ h 3.4.2 Pool Liner ! The stainless steel pool liner shall be removed following the removal of the pool internals. The liner will be torch cut or mechanically cut and sized for decontan.ination and/or disposal. If stainic$s steel is torch cut, air !
- a supplied respirators will be required because of generation of nickel carbonyl.
During the early efforts at alleviating leakage- from the-reactor pool, fiberglass materials were used on the inside walls of the pool, No reference-is made in any of the available documentation to removing the fiberglass ; materials prior to installing the stainless steel liner. Therefore, it should be assumed that the fiberglass liner material remains between the stainless ; steel and the concrete walls of the pool. Extra respirator requirements will ; be enforced during removal of the stainless steel liner to prevent the ! inhalation of fibers during cutting operations. Should the fiberglass be on the pool walls, this task will also include removal and disposal of such , fiberglass materials. 3.4.3 Platforms i i The platforms surrounding the reactor pool monolith shall be removed af ter > removing the pool internals. A cross section of the reactor building is- shown ( l- in Figure 3-4. The first step in this phase shall be the removal of insulating materials I and lighting fixtures from the interier of the ellipsoidal head of the containment. The removals will be' accomplished by constructing scaffolding on !
'. the bridge crane to gain access to the underside of the head. Access to the crane is by way of the roof of'the control room. The next step would be the L
?.
removal ~of the control room structure and related equipment on the second .; platform shown in Figure 3-5. Care must be taken to ensure that electrical power supplying the areas being worked on have been deenergized prior to removing electrical equipment and fixtures, Remote operating-tcols, stored on , 3 15 _ _ . _ . _ . - . , . . . , . . . . . _ _ . _ . . , _ . , _ _ _ . - _ . _ - . . _ _ _ . _ , _ . _ - _ . . _ . _ . _ . ~ _ _ . . _ _ . _ _ . . . _ . . . _ . . _ _ . . _ _ _ _
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__ ___ _ - . _ . . _ . _ - _ _ _ . = - _ . _ _ _ _ _ _ . _ _ _ , _ - a the second platform railing, and the HVAC equipment and ducting will be removed as part of this task. The stairway connecting the first and second platforms will be removed prior to removing the second platform decking. As HVAC and ^ H <t r p aquipment are removed, temporary ventilation, . electrical power, Lno Abe must be provided to ensure a safe working environment durin, <isa v4 -apu ig. The temporary services will also include emergency lighting (a tea asen+ of a power failure. l-1he decking of the second platform will be demolished using standard , industry accepted techniddes, Care must be taken when working in the area adjactnt to the pool monolith and the containment wall so that these -l structures are not damaged until they are scheduled for removal. 1hc supports for the second platform will also be removed at this time. i MQlEl At this point, the contractor may elect to integrate the demolition of the pool wits the removal of the platforms. That is, after the removal of the second platform, the contractor could initiate the demolition of the_ reactor pool _ structuro down to the level of the first platform then remove the first platform-ano another section of pool structure continuing down to the basement level in a similar manner. The first platform, shown in Figure 3-6, will be removed in a similar manner as'the second platform, as explained above. The conduits, piping, ductwork, and stairway will be removed first, followed by removal of the decking and supports down to the level of the main floor. 3,4,4 _ Basement Electrical - At the point in the project where all planned usage of the installed . electrical system has passed, the electrical distribution system located in the basement can be-deenergized_and the equipment removed. This removal will ' Linclude the motor control centers, distribution boxes, conduit, and other , installed equipment. 3-18
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i Caution should be exercised to ensure that all circuits are deenergized, ' including shorting out all capacitors prior to starting removal activities. 3.4.5 Basement Piping and Equipment 4 The basement contains llVAC, water conditioning equipment, coolant piping - and components, barricades, and other miscellaneous equipment. These items , will be removed from the facility in a convenient order, with the coolant .. related equipment being removed last to minimite the potential for l contamination of materials. As in dismantling other piping systems, steps i; must be taken to ensure that systems are empty and not pressurized. i The' basement equipment can be hoisted to the main floor through either of i j- the 7 by 7 ft removable hatches in the main floor (see figure 3 7).
.i The coolant equipment enclosure walls (figure 3-8) should be removed prior to removing equipment in order to facilitate access to the equipment.
- 3.4.6 Basement Sumps i The two basement dry sumps (figure 3-8) shall be removed by cutting the steel liners away from the concrete. The concrete will then be surveyed and I (hipped away as needed to remove contaminated materials.
The floor drains and connecting lines shall be chipped out from the drain inlet to where the lines enter the sump. ' j The sump pump shall be removed, the sump liner cut away from the ; concrete, and concrete removed, as needed, and materials disposed of as , surveys and analyses dictate. ! i Openings made into the floor system shall be covered and/or barricaded to preveni' personnel injury from falls. [ l. 3- 'O .
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. . -- . = _ . _ . _ . _ _ . _ . . __ . _ . . - . . _ . _ _ . . _ -
j 3.4.7 Gama facility and Storage Tubes 1he gamma-ray facility (figure 3 8) steel liner and storage rack shall be f ] cut away from the concrete and the concrete surveyed and scabbled as needed to f,. remove contamination. The storage tubes (figure 3-8) shall be removed ising i coring, cutting, or other methods chosen by the contractor to tieparate the tubes individually or in groups from the basement floor. The openings resulting from removal of the storage tubes shall be temporarily covered to prevent personnel injuries. 3.4.8 Pool if not phased in earlier, the pool demolition is the next step after the basement has been decontaminated and all materials removed, if pool demolition was begun earlier, progress can now be made beyond the main floor. During pool demolition, the operating floor shall also be removed back to the supporting columns. pool structural materials are very likely to be contaminated in the area above the reactor core and for a short distance below the core area, lhere is alsu a high probability of contamination in the hold-up tank below the reactor pool. Extra care, such as the use of temporary work enclosures, must be taken during this phase to avoid unnecessary exposure to wo>kers and to prevent the spread of contamination. Demolition of the pool monolith can be accomplished utilizing chipping techniques, sawing, drilling and wedgin s or other methods proposed by the contractor and approved by the Army. In any case, the methodology to be used in this phase must be approved by the Army before implementation to ensure that the method will minimize the potential for spread cf airborne contamination. 3-?3
Approximately 51 ft of lead is ptesent in the lining of the floor 3 storage pits and boxes in the annulus as well as between the horizontal beain tubes and the pool wall. This lead will be surveyed for radioactive contamination. Cont aminated lead will be packagad and delivered to the licensee for storage as mixed waste, Uncont"ninated lead iill be salvaged for , reuse. If the lead has removable contamination, the lead shall be decontaminated and salvaged for reuse. ,, Specific attention shall be given to checking the rebar removed f rc.m the pool structure for the presence of contamination and/or activation. 3.4.9 Building Internal Decontamination following removal of building internal components as previously described, all that will remain will be the A.ilding shell, the concrete wall, basement floor, operating floor f rom the wall to the supporting columns, and the airlock between the two buildings. A survey of the building interior shall be performed, and if radioactivity above releasable limits as defined in Reference 4 is detected during this survey, decontamination efforts, including scabbling, will be used to remove the contamination, This survey and subsequent decontamination (if decontamination is required) are necessary to ensure that there is no residual radioactive contamination above releasable limits. The survey of the building interior will be performed in accordance with a written procedure prepared by the decommissioning contractor and approved by the licensee. The precedure will be based on the radielogical contamination encountered during the performance of tasks described in Chapter 3, Section 3.4, and the procedure will reflect the applicable parts of Reference 2. 3.5 TERMINATION SURVEY lhe termination survey / characterization methodology is described in detail in Chapter 8 but is briefly summarized here. The termination survey will consist of two phases, separated by backfilling and grading the trenches 3-24
and pits formeu when piping and tanks are excavated. Each phase will be performed by the decommissioning contractor with a follow-on verification survey to be performed by the U.S. Ar my Environm(ntal liygiene Agency (USA [HA). 3.5.1 Phase 1 At the conclusion cf removing Cistern 242, the secondary coelant system, and associated / piping, there will be sever &l trenthes and holes in the terrain around the reactor building. ihese excavated areas shall be radiologically characterized to demonstrate tc the NRC that the residual contanination levels meet NRC release criteria. This phase of the termination survey must be performed prior to backfilling the holes created by excavation in order to prevent covering potential contamination with several feet of soil. In addition to soil characterization during Phase 1 of the termination survey, the entire inside surface of the reactor building shall be gridded and characterized. This characterization will consist of direct ,4adiat'9n measurements and analysis of smears and other samples collected after completion of decontamination as described in Section 3.4. 3.5.2 Phase 11 After backfilling the excavated trenches and holes and grading the site, Phase 11 of the termination survey shall be performed as specified in Chapter 8. 3.6 BACKFILL AND GRADE Following approval of Phase I of the final survey, the trenches and holes caused by excavation will be backfilled and graded. Backfill soil will consist of excavated soil. 3-25 It _ - - - - - - - _ - - - - _ - - _ _ _ _ _ _ _ _
3.7 ESTIMATED SCHEDULE AND COST Figure 3-9 is a critical path method (CPM) network showing the major project activities to be performed. These include activities to be performed by the Army as well as activities to be performed by the decornissioning , contractor. The CPM network shows requirea sequencing of activities and also shows activities that may be performed in parcllel. Estimated duration for c acti'vities is shown in weeks above eacn activity line. 5 The estim- cost of the project is $5.1 million assuming a radioactive waste void v' of f.0%, and 54.3 mi; ion assuming a void /olume af 10%. The actual void volume will depend on the efforts by the decommissioning contractor to minimize void volume. A breakdown of this estimate is given in L Table 3-1. The cest estimates are based on partial dismantlement, variaticns 1-B and 1-D, described in Reference 1. These variations assume a void volum of 10% and 60% respectively in the packaged radioactive waste. Also, it is assued that the radioactive waste would be disposed of after 1 January 1992, but before 1 January 1993. Disposal of the waste during that period requires a state penalty surcharge cf $120 per cubic foot for low-level radioactive waste. 3.8 FUNDING Funding for decommissioning the AMTL Reactor will come from appropriate U.S. Army sources. The clcsure action at AMTl. and the tui,d; required for this action are mandated under Public Law 100-528. -
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9 3-26 I
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CPM Network for Decommissioning the AMTL Reactor Terminate building 100 operations A Remove equipment MTL MTL (15) (10) Select decommissioning (2) (22) Prepare SOW contractor p Mobilize resources Perform decommissioning (p01Submit
, DP to NRC r 05 r 10 r 15 r- 20 25 v Licensee COE-NED COE-NED & MTL Contractor Contractor rk (26)
Review of DP ano NEPA (1) 1f documentation issue decommissioning order 07 Y NRC NRC C A A 7 7 { , (B) (4) (1) (8) Perform Phase i Evah'a4 Backfili Perform Phase !! Evaluate Phase il results and termination survev request release of facility Phase I resu'ts trenches & pits p termination survey ,- Contractor /USAEHA MTL & NRC Contractor ContractcfAJSAEHA Licensee Note: Number in parentheses above each activ:ty line is estimated duration in weeks T91 c505 Figure 3-9. Critical Path Method (CPM) network for the AMTL Reactor decommissioning.
Tabic 3-1. Cost estimate summary for decommissioning the AMIL Reactor. Estimated Costs Pro.iect Activity and WES Numbfr ,_, JM)* JEf Site Preparation 3.1 6* 6*
-Project Management 3.2 378 378 -
Auxiliary Structures ne.cval 3.3 25 25 d Reactor Bldg. Decontamination
. 3.4 3300 3994' "
Termination _ Survey 3.5 108 108 Backfill and Grade 3,6 22 22 Subtotal 3839' 4533' Contingency (13%) _422_ _q22_ Total 4338 5122
- a. This cost estimate is based on 10% void volume for radioactive waste,
- b. This cost estimate is based on 60% void volume for radioactive waste.
- c. The cost estimate for site preparation covers only the. tasks to be performed by the decommissioning contractor described in Section 3.1, and does not cover tasks to be performed by the Army,
- d. -This cost estimate includes $960K for the-state penalty surcharge for the 4
disposal of 8,000 cubic . feet of low-level radioactive waste generated in Massachusetts,
- e. This _ cost estimate includes 51440K for the- state penalty r. 'harge for the 1:posal of 12,000 cubic feet of low-level radioactive v 2 enerated in Massachusetts,
- f. Subtotal includes $422K of escalation to uan activity miugou~ of September
'1992.
D L l 3-28
CHAPTER 4 SECURITY I
- 4. INil10 DUCTION The nuclear fuel was previously removed from the AMTL Reactor and shipped off site. Therefore, there is no requirement for safeguarding special nuclear material.
I 4.1 PHYSICAL SECURITY Ouring decontamination of the AMIL Reactor as specified in Section 3.4, access to the inside of the reactor building will be controlled, using the personnel entrance between Building 97 and the reactor building. An appropriate barrier will be installed in the airlock. Equipitent and material will be moved in and out of the reactor building using the double door shown in Figure 3-7. During decommissioning as specified in Section 3, access to the site will be controlled using standard construction practices such as warning signs, fencing off the area, and installing physical barricades around excavations. Access will be limited to the construction site as required by the NRC for radiation areas, high radiation areas, and areas containing radioactive material as specified in 10 CFR 20. 9 4 4-1
l CHAPTER 5 RADIOLOGICAL ACCIDENT ANALYSIS A Radiological Accident Analysis is not required for this DP because there is no nuclear fuel at the AMIL Reactor site. Potential radiological accidents, during decommissioning activities of the facility, are discussed in
. Sections 3.3 and 3.4.
9 5-1
, _ _ ~ .
~
CHAPTER 6 RADI0 ACTIVE MATERIALS AND WASTE MANAGEMENT
- 6. INTRODUCTION During deconunissioning activities, radioactive wastes in liquid, solid, and particulate forms are expected to be generated. Planning the management of these wastes is an integral part of the DP. Provisions for minimizing the amount of waste generated, and waste collection, treatment, packaging, and ;
shipment off site for disposal are discussed in the following sections. l 6.1 FUEL DISPOSAL The fuel was removed from the AMIL Reactor during deactivation in the early 1970s. l 1 6.2 LIouro RADIOACTIVE WASTE Liquid radioactive wastes generated during decommissioning activities l will be collected, monitored, and released to the sanitary sewage system if the conditions of 10 CFR 20.303 can be met. The project manager will contact the Massachusetts Department of Environmental Protection (DEP), Massachusetts Water Resource Authority (MWRA), and the NRC in order to obtain all permits
- and certificates required to discharge these wastes into the local sanitary or storm sewer systems. Contaminated water will be treated to remove radioactive l contamination, or processed for disposal as low-level radioactive waste. . Discharge shall be in accordance with 10 CFR 20 and the DEP and MWRA for waste water. Discharge must be approved by HQ AMCCOM. ~
Possible sources of liquid radioactive waste are:
. Decontamination of components and parts 6-1
- Personnel decontamination liquids
- Decontamination of structures and floors
- Residual liquid in piping and other componcnts. ,
Efforts will be made throughout dec3mmissioning activities to minimize ,, generation of liquid waste, Whenever possible, scrubbing with swabs will be used instead of spraying. The water mist used during the demolition of activated concrete will be closely controlled, and approved liquid absorbers will be used around the floor to absorb any runoffs. Liquid waste will be absorbed or solidified using absorbent materials or solidification agents specified in the license requirements of the radioactive waste disposal site. Absorbent material will be provided to absorb at least twice the volume of radioactive liquid contents in all radioactive waste packages. 6.3 Sotto RADIOACTIVE WASTE The solid radioactive wastes generated during decommissioning activities will be packaged on site in containers suitable for shipping and disposal at an NRC approved disposal site in the United States. Packaging requirements shall be in accordance with the latest Federal Regulations, and the packaging requirements shall also comply with the requirements of the NRC approved disposal site chosen. 6.3.1 Packaging , The types of solid radioactive waste to be packaged include the following: Demolition Materials--These include all contaminated and/or activated systems, components, and equipment removed during 6-2 1 l I
dismantlement of the AMIL Reactor facility described in Section 1.1.1. Packaging these materials will be specified in work procedures written by the contractor and approved by the Army.
. . Equipment and Tools--This includes such items as saws, jackhammers, forklift, shovels, pumps, tanks, ventilation system components, filters, and piping. Not all of this equipment is expected to be discarded as radioactive waste. A determination of volumes of solid radioactive waste generated from this category will be possible only during cleanup, when measurement of contamination level and evaluation for decontamination will be made.
. Auxiliary Materials and Clothing--This includes confinement barrier plastic sheets, protective mats, rags, work platforms, and protective clothing. 't is assumed that these materials will be compacted and packaged in 55-gal drums if contaminated.
6,3.2 Temporary Storage of Radioactive Waste
^
Containerized waste shall be transported to the chosen disposal site in a timely manner to avoid accumulation of waste. However, in the event that packaged waste must be temporarily stored, it will be stored in a protected , area near the reactor facility. 6.4 VENTILATION SYSTEM Ventilation exhaust systems will be equipped with roughing filters to
., capture large particles, and with high-efficiency particulate absorption (HEPA) filters to provide up to 99.99% particalate retention. HEPA filters will be changed in the event of high radiatica level readings / alarm in the exhaust duct, or based on maximum pressure differential readings indicating that the filters are filled with dust. Radiation monitoring will be provided for the exhaust air to atmosphere; readings above prescribed limits will shut off the exhaust fans. Ventilation unit. serving confined enclosures will be 6-3
mobile type and connected with flexible ducts to the enclosure and to the exhaust duct. The exhaust duct will be provided with gravity louvers, which will automatically close in the event of f ailure of the ventilation exhaust unit. 6,5 WASTE CLASSIFICATION The criteria for waste classification for low-level waste disposal are contained in 10 CFR 61. -- It is concluded that most all the radioactive wastes frcm this project can be classified as Class A unstable waste, and will be segregated at the disposal facility. Items that can be compacted shall be sent to the Defense Consolidation Facility. 6.6 SHIPPING RADIOACTIVE WASTES Unless otherwise specified, it is assumed that most shipments will be low-specific activity (LSA) and will be shipped in exclusive use vehicles. Department of Transportation (DOT) Regulation 49 CFR 173 provides _ radiation level limitations for transportation of packages of radioactive materials in closed, exclusive-use transport vehicles as follows: External radiation levels must not exceed 1000 mrem /h on the accessible surface of the package, if the shipment is made in a closed transport vehicle, the package are secured, and there is no . loading or unloading operations during transit (49 CFR 173.441) '
. The radiation level must not exceed 200 mrem /h at any point on the -
outer surface of the transport vehicle, 10 mrem /h two meters from the vehicle sides and 2 mrem /h in the tractor cab (49 CFR 173.441). 6-4
A quality control program shall be established by the decommissioning contractor to ensure that radioactive wasie shipping regulations are enforced. The decommissioning contractor shall have the appropriate DOT certification. The decommissioning contractor's quality control program shall include the
.' following requirements:
.. . Waste Containers All Type A containers or Type B containers if the A quantity is exceeded are D01 specification 7A (49 CFR 173.425, 178.350, and 173.24). Exemptions to this must be specified. Most containers for this project will be strong, tight containers, not Type A or B. This assumes LSA exclusive use vehicle shipments.
All containers are in good physical condition, with no evidence of damage, corrosion, or leakage (49 CFR 173.475 and 173.24). All metal drums with a capacity of 55-gal or greater will have a 5/8-in. or larger bolt for securing the closure device (ring assembly). All metal containers will have an intact heavy-duty closure device with gasket when presented for disposal. Ring bolt is torqued to approximately 45 ft pounds (recommended). Drum lids are a proper fit, and bungs (if any) are tight
., (49 CFR 173.475).
Radiation levels at the package surface are limited to those specified in 49 CFR 172.403 for packages requiri19 Yellow-ll or White-I labels and to those specified in 49 CFR 173.441 for all other packages. 6-5
I Surface contamination levels are below DOT limits (49 CFR 173.443) 220 alpha 'dpm/100 cm2 and 2200 beta-gamma
-dpm/100 cm'.
- Labels and Markings (if not excluded because of exclusive-use vehicle exceptions) -
Each container ha tive labels affixed to it (49 CFR 172.403 L 15). Each container has b. - .
'tive Material, LSA, n.o.s. UN 2912" in great .high letters (49 CFR 112.101, 172.301).
Each container has been marked " USA DOT 7A, Type A" or
" DOT. Type B" in 1/2-in.-high letters (49 CFR 178.350 and 172.310),
The-waste class has been-marked on each container using greater than 1/2-in.-high letters (10 CFR 61.57). Waste Class Exam _ple Class A Unstable Class A Stable , Class B Class C i
'For each container in excess of 110 lbs, the weight and unit of- , ,
measurement have been marked on the container (49 CFR 172.'310). Markings must be durable and legible, and displayed on a ~ background of sharply contrasting color, unobscured, and located away-from any other marking, such as advertising, (49 CFR 172.304). 6-6
- . - -. a-
I
- The name and address of the shipper has been attached to each container if vehicular transfers are involved (49 CFR 172,306).
- Transport Vehicle The total transport index number does not exceed 50
.. (49 CFR 177.842). This is not applicable to exclusive-use vehicle shipments.
- The containers have been loaaed, blocked, and braced so that they cannot change position during conditions normally incident to transportation (49 CFR 177.425 and 177.842d).
. Exclusive-Use Vehicle Exceptions Specific instructions for maintenance of exclusive-use shipment controls must be provided by the shipper to the carrier and be included with the shipping paper information (49 CFR 173.425 b9).
Ar exclusive-use vehicle shipment of Type A low Specific Activity (LSA) radioactive material (RAM) is exempt from Type A packaging specifications if it meets a strong, tight packaging criterion (49 CFR 173.24). Shipments must be loaded by consignor and unloaded by consignee. Tha consignor shall be the U.S. Army Armament,
., Munitions, and Chemical Command (AMCCOM) representative who has the authority to approve packaging and shipping radioactive waste from this project. The AMCCOM representative will be onsite full time.
There must be no loose radioactive material in the conveyance. 6-7
- Shipment must be braceo so as to prevent shif ting of' lading ,
under conditions normally incident to transportation ] (49CFR173.425), For shipments of LSA radioactive materials or shipments ' containing packages bearing Radioactive Yellow-Ill labels, the transport vehicle shall be- placarded in accordance with Table I .. of 49 CFR 172.504. Exclusive-use vehicle shipments of LSA RAM are exempt from specified markings and labeling if the exterior of each package ., is stenciled or otherwise marked " Radioactive - LSA." All-DOT regulations listed in this section are current regulations. The most current, applicable regulations shall be used when-the shipments are made. -
.- Documentation
-The generator has a valid disposal-site and (where required) state
- User' Permit,
.All required documents are fully completed and are legible.
The Radioactive Shipment Manifest (RSM),'formerly RSR/ Manifest, is complete-in all details,- (49 CFR 172.200 Subpart C, 10 CFR 20.311, and radioactive wasta disposal site requirements.) ,
- The number of containers listed on 'the RSM agrees with the physical count of containers loaded, m -
All required certifications, RSMs, and other documents as appropriate are signed. L 6-8
The use of abbreviations must conform to DOT and NRC specifications.
. Shipping Routes The route cut 4MTL to the NRC approved disposal f acility shall be chosen using tne PC Miler Program, which calculates the shortest highway routes. Local, state, and federal ordinances may impact the routes selected to the chosen disposal site. Currently, the cost estimate is based on shipping the radioactive waste.
If shipments are made to Barnwell, prior notification must be sent te the State of South Carolina for each shipment. The prior notification must be received at least 72 hours before the shipment arrives at the Barnwell disposal facility. e 6-9
CHAPTER 7 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS
- 7. INTRODUCTION Technical and environmental specifications will be implemented by the decommissioning contractor to control conditions, parameters, and variables so
- that during decommissioning activities the radiation exposure to workers and the public shall be maintained as low as reasonably achievable.
The technical and environmental specifications will include items in the following categories:
- Health and safety limits
- Surveillance requirements
- Administrative controls a Design features.
7.1 HEALTH AND SAFETY LIMITS Radiation and industrial exposures to workers and the public must be ALARA and less than regulatory limits. Under no circumstances will the exposures exceed regulatory limits as specified in 10 CFR 20. During decommissioning activities the following limits shall be enforced. 7.1.1 External Exposure External radiation exposure for individuals in restricted areas during decommissioning shall not exceed the limits specified in 10 CFR 20.101. 7-1
__ _ .m. .-_m ___ - __. __ . _ . _ _ _ _ _ . . _ _. _ _ _ _ - _ _ _ . 7.1.2 Internal Exposure Internal radiation exposure from Inhalation of radioactive material in air in restricted areas shall not exceed that which would result from the inhalation of the limiting quantities specified in 10 CFR 20.103, lhis will be verified by performing a baseline bioassay prior to assigning contract ,
~
employees to the restricted area. After the decommissioning has been completed, a termination bioassay will be performed for all individuals potentially exposed to airborne radioactivity. 7.1.3 Concentration of Airborne Radioactive Materia'. in Restricted Areas Concentration of airborne radioactive material in restricted areas shall
,ot exceed the limits set in 10 CFR 20.103, 7.1.4 Concentration of Airborne Radioactive Material in Unrestricted Areas Concentration of airborne radioactive material in unrestricted areas shall not exceed the limits specified in 10 CFR 20.106.
7.1.5 Concentration of Nonradioactive Substances in Restricted Areas Concentrations of nonradioactive substances in restricted areas shall not exceed limits established by the American Conference of Governmental Industrial Hygienists (ACGlH) as listed in " Threshold Limit Values (TLV) and Biological Exposure Indices for 1987-1988." l 7.1.6 Concentration of Nonradioactive-Substances in Unrestricted Areas Concentration of nonradioactive substances in unrestricted areas shall -
-not exceed 1/30 the ACGIH TLV limits.
7.1.7 Noise Levels Noise emissions shall be in compliance with Title 310 of Code of Massachusetts Regulation 310 CMR 7.10 and Town of WaterTown Ordinances 7-2 l
Chapter Vlil, Section 32. Protection against the effects of noise exposure to workers shall be in compliance with U.S. Army Corps of Engineers Safety and Health Manual, EM 385-1-1. 7.1.8 As low as Reasonably Achievable (ALARA) The AMTL Commander is committed to the concept of ALARA in terms of both individual and collective doses. Administrative controls, training, protective devices, and mea ures will be used to achieve this commitment to ALARA and minimize the hazards to health and safety. The ALARA pronram shall comply with ASTM E 1167, Standard Guide for Radiation Protection Program for Decommissioning Operations. 7.2 SURVEILLANCE REQUIREMENTS Surveillance activities to be conducted during decommissioning are described below. 7.2.1 The Dosimeter Program A dosimetry program will be developed and implemented by the decommissioning contractor in compliance with state and Federal requirements. Vendors supplying dosimeters shall have NAVLAP accreditation in compliance with the 10 CFR 20. 7,2.2 The Routine Swipe Program All material, components, and equipment remcved during decontamination will be swiped to determine their disposition relative to removable contamination. The swipe program will be specified in procedures prepared by the contractor, and must meet measurement requirements to detect removable surf ace con'.amir sn at the most restrictive levels specified in Reference 4. 7-3
_ _ __ __. .._ _ _ _ . - . _ _ . _ - . _ - ~ _ _ _ _ _ _ _ _ _ 7.2.3 The Routine Instrument Survey Program Direct-radiatio: measurements will also be performed on all material, components, and equipment removed during decontamination to determine their disposition relative to fixed contamination or activation. Details of the survey program will be specified in procedures prepared by the contractor and , must meet measurement requirements to detect surface contamination at the most restrictive levels 1,pecified in Reference 4. .. 7.2.4 The Air Sampling and Monitoring Frogram Continuous air monitors will be located outside the reactor enclosure. Monitoring inside the enclosure will be conducted for personnel exposure under the Health and Safety Program. Details of this program will also be specified in procedures prepared by the contractor. The procedures shall comply with ASTM E 1167, Standard Guide for Radiation Protection Program for Decommissioning Operations. All instruments / systems used for monitoring purposes shall be properly l calibrated prior to use in compliance with ANSI N 323, Radiation Protection l Instrumentation Test and Calibration. 7.3 ADMINISTRATIVE CONTROLS The administrative controls that will be used during decommissiening operations are briefly discussed below as are the responsible organization and I documentation requirements. 7.3.1 Administrative Controls During Decontamination , Administrative controls during decontamination are the provisions relating to organization and management, proceduces, record keeping, review and audit, and reporting necessary to ensure completion of the decommissioning in a safe manner. Administrative controls shall be in Compliance with ASTM E 1167. 7-4
7.3.2 Responsibility The licensee shall have overall responsibility for the completion of decommissioning of the AMIL Reactor. However, the contractor's Project Manager will be delegated the responsibility for decommissioning via the contract. 7.3.3 Organization The organizational structure for management and performance of the det.ommissioning activities is shown in Figure 1-22. The functions and responsibilities, ano minimum required qualifications and experience of each key position are detailed in Section 1.9. Qualifications of decommissioning personnel shall comply with ANSI 3.1, Standard for Selection, Qualification, and Training of Personnel for Nuclear Power Plants. 7.3.4 Recorcs and Reports Accurate and complete records and reports shall be maintained by the contractor in compliance with 10 CFR 20 sid other applicable regulations. The records and reports will cover performance and completion of all activities that may have resulted in exposure of workers or the public to radiation or other hazardous / toxic materials.
+ Records Records shall be maintained in compliance with ANSI 15.10. Records f to be maintained during decommissioning shall include those listed , below.
L- Health and Safety Related Activities: i
- Work permits Work procedures Radiation survey reports Contamination survey reports 7-5
Airborne survey reports
- Environmental survey reports
]
Counting data on air samples, smears, and gamma spectrum I analysis Instrument calibrations 1 Source inventory and storage .!
- Radioactive material inventory and storaga ;
Shipment records .. ; Waste disposal - surveys and records Packaoe cortifh.ations/ records l Incidents and accidents
- Confined space entry permits Continuous monitoring records for oxygen deficient / explosive atmospheres, Personnel Rg_c_qnf1
- Bioassay analysis Personnel exposure records Individual dosimeter readings as related to daily tasks and work procedures Respiratory protection qualifications (medical clearance and fit test)
Audiogram results
. Training records Visitor logs and exposure information.
. Reports
. 4 Reports prepared by the decommissioning contractor shall be written and submitted to the US Army COE-NED Contracting Officers ,
representative. 7-6
. Review Responsibility for review of procedures, practices, and performance shall rest with the appropriate individuals anofor committees specified in Section 1.5.
7.4 ENGINEERING CONTROLS o. Confinernent barriers and HEPA filtered ventilation systems will be employed during decontamination in Compliance with ASTM E 1167. These systems will provide:
- Maintenance of a negative pressure in confined areas. This will be accomplished by a HEPA filter system in conjunction with the installed confinement barrier.
. Provisions for dust control during decontamination to minimize airborne particulates.
- Provisions for air monitors at outlet of ventilation systems to turn off ventilation system if excessive levels are measured.
-9 O 7-7
CHAPTER 8 PROPOSED TERMINATION RADIATION SURVEY PLAN
- 8. INTRODUCTION The termination radiation survey (characterization) of the AMIL Reactor
.. and site will be performed in two phases. Each phase will require both the decommissioning contractor and the U.S. Army Environmental Hygiene Agency (USAEHA) to conduct independent surveys. lhe contractor's survey will be conducted to ensure that no residual radioactivity above releasable levels remains at the facility or site. The USAEHA survey, which will be conducted subsequent to the contractor's survey, will serve as the means of validating the conclusion of the contractor's results. The first phase of the termination survey will include the reactor building interior and the trenches and pits formed during required excavation outside the reactor building, lhe required excavation is described in Chapter 3. The second phase of characterization will consist of the soil area outside the reactor building and will be performed following backfilling and grading the excavated regions of the site.
This chapter addresses background soil and the two phases of characterization to be performed after partial decontamination. The surveys performed on removed components and material for disposal were described in Chapter 3 and are not included in this chapter. The determination of radiological background for this project and the performance of each phase of the termination radiation survey will be in
. accordance with written procedures. The decommissioning contractor and the USAEHA will use acceptable procedures which will be submitted to the licensee for approval.
The decommisioning contractor's procedures will be in accordance with the gene al criteria of Reference 2 and the specific requirements specified in this chapter. The USAEHA procedures will be based on the verification 8-1
inspection criteria of Reference 2. In addition, both sets of procedures will establish data quality objectives to be met by each performing organization,
8.1 BACKGROUND
SOIL Results of soil. measurements and analyses during both phases of the termination radiation survey shall be compared with background radiation , levels to determine subsequent actions relative to the soil-areas. Rel iable background 5011' data are essential and must be collected prior to performing the measurements and analyses of the soil. Features that contribute to good background data are (1) 3urvey design to provide representative, unbiased sampling, (2) proper allocation of sampling, (3) selection of areas least likely to have been affected by the reactor or other AMTL operations, (4) appropriate instrumentation, and (5) quality ensured analyses. Soil background locations for this project will be determined by drawing twelve radius lines _ from the _ center of the reactor, and extending each radius line out to 3.5 km from the reactor center. Each of the twelve radii will be 30 degrees apart. Next, draw concentric circles with radil of 0.5, 1.5_, and 3.0 km, with the center of each circle at the reactor center. The background sample locations will be selected at the intersection of each radins line and concentric circle. If there are obstructions or other reasons preventing accessing soil at those locations, adjust the distance or angle in order to find accessible soil. Perform a soil contact radiation measurement of alpha, beta,_and , beta-gamma at each of the 36 locations. Record each measurement in 2 dpm/100 cm2 . Recording the measurement in dpm/100 cm requires converting to 2 dpm/100 cm by-knowing the instrument efficiency and the active area of the detector. i I l 8-2 I
l In- addition, collect a sanaple f rom the top 15 cm of soll at each location ! for analyses. The analyses shall include gross alpha and beta concentrations. ; Also, ~ gamma spectrometry shall be performed for each sample, and the # concentiation determined for each detectable radionuclide, , ,.- 8.2 PHASE I
... 8.2.1 Reactor Building Interior Survey All interior surfaces remaining in the reactor building af ter completion of decommissioning described in Chapter 3 shall be surveyed for fixed and removable alpha and beta-gamma residual contamination.
The interior of the reactor building following decommissioning will consist of the basement floor, the basement walls, the' operating floor from
- the reactor walls. out- to the supporting columns (the center portion of the operating floor will have been removed during decimmissioning), the reactor building walls, and the ellipsoidal ceiling. No neutron activated material will remain in the reactor building following decommissioning. Therefore, no ,
concern need be given to subsurface activation products.during the termination survey. Two degrees of survey intensity shall be used, level I and Level- II. Level _I will be the more survey-intensive level and will be performed in the areas most likely to contain residual. contamination such as the basement floor, sumps, anu pits. Levei 11 intensity surveys will be performed in areas less likely- to contain residual radioactivity such as the reactor walls above the operating floor and the ceiling. The areas to be designated Level I and Level 11 Elli be determined after completion of decommissioning activities and will be based on the contamination discovered during decommissioning. Level'I grid size shall be 1 meter X 1 meter and Level 11 grid size shall be 3 meter X 3 meter. The details of the Level 1 and 11 surveys will be specified in' procedures prepared by the decommissioning contractor. The procedures prepared by the 8-3
,. - - - - _ ... _ -. .. - ~- - - -. - - .. . ...- - - -~ - .-. .. . .-.
b USAEHA will use-the same Level I and 11 designated areas as the e decommissioning, contractor. Both sets of procedures will be approved by the licensee. The objective of the surveys of the AMit Reactor building interior is to show that residual fixed or removable alpha or beta gamma contamination is ,
~
below acceptabl_e' levels specified in Reference 4. - The survey-procedures shall incorporate principles described in
=NUREG/CR-2082 and include the following requirements: ,
. The level I and 11 surveys shall include direct radiation measurements to determine alpha and beta-gamma surface contamination in. disintegrations per minute (dpm) per 100 cm2,
. Smears sSall be collected and analyzed to determine removable alpha and beta g uma surface contamination in dpm per 100 cm2 .
.. Measurements must be performed using instruments capable of measuring alpha and' beta-gamma radioactivity below releasable levels -
specified in Reference 4 for applicable radionuclides. If radionuclides are not known, the most restrictive levels in Table I _ of Reference 4 shall be used.
.- In addition to measurements to determine surface fixed and removabl.e
! . contamination, exposure rate measurements shall be performed in accordance with Reference 4. 8.2.2_ Radiological Characterization of Trenches and Pits i
--Figure G-l?is a sketch of the AMIL Reactor site after excavation but-prior to backfilling and grading.s The excavated areas with potential
- contamination are the cistern pit, the pit where the secondary coolant system sump was previously located, and the piping trenches. In addition, the berms
.of excavated soil may contain contamination.
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The bottoms of the two excavated pits and trenches will be grided into 1- x l m grid squares. The berms of excavated soil will aise be grided into 1- x l-m grid squares, lhe Phase I characterization will include soil contact radiation measurements for beta-gamma at each grid intersection and at the center of each grid square. lhe measurement will be performed t'y integrating counts for , at least 30 seconds, converting the measurement, and recording in dpm/100 cm . Dhase I characterization will also include collection of surface soil samples from 50% of all grid squares (randomly selected) in the bottom of each pit and trench. In addition, a soil sample will be collected at depths of I ft and 3 ft from 50% of all gnid squares (randomly selected) over the berms of excavated soil. These soil sbmples should adequately represent the excavated soil, pits, anJ trenches because of the homogenization that takes place during excavation. Each soil sample will be analyzed for gross alpha and beta concentration and gamraa spectromelry. If gross alpha or beta is detected statistically significant above background, the analytical results of that sample will be verified. If results are positively verified, an appropriate alpha or beta isotopic concentration analysis will be performed for each sample exceeding gross alpha or beta in background soil. 8.2.3 Evaluation of Phase 1 Results The results of the decommissioning contractor's Phase I survey, including soil analyses, will be summarized in a Phase 1 Characterization Report. This report will be submitted in Jraft form to the USAEHA for review. After completing their review, the USAEHA will attach the results af their own Phase , I survey as an addendum to the contractos results and the report will be submitted to the licensee.. Backfillint will not be performed until authorized . by the licensee. The licensee's decisi'n will be based on the evaluation of the Phase I soil survey results. If either evaluation reveals contaminated soil above releasable levels, the contaminated soil will be removed and that area resurveyed, using the same methodology as specified abnve. If either 8-6
l evaluation reveals surface contamination inside-the reactor building above acceptable levels, additional decontamination shall be performed ano those areas resurveyed. During the time excavations are open, the licensee will notify the NRC and invite them to perform a confirmatory, independent radiological characterization of the soil prior to backfilling. During the period of time between the completion of excavation and the finalization of the survey
results, the excavations will be protected from the intrusion of precipitation and the dispersion of soil particles due to wind erosion, 8.3 PHASE 11 Phase 11 characterization will begin after the pits and trenches have been backfilled and graded.
s The backfilleo and graded areas will be grided into 1- x l-m grid squares. Beta-gamma soil contact radiation measurements shall be performed at each grid intersection and at the center of each grid square. Each contact measurement shall be made by integrating counts for at least 30 seconds, - converting the measurement and recording in dpm/100 cm2 . 4 In addition,-a gamma measurement will be made at 1 meter above the-soil surface at each grid intersection and the center of each grid square. This measurement will bc recorded-in pR/h. Surface and subsurface soil samples shall be collected in the center of 50% of all the-grid squares (randomly selected). Each surface sample shall be collected from the top 15 cm of soil. Each subsurface sample shall be collected at depths of five, ten, and fifteen feet.
- Each soil sample will be analyzed for gross alpba and beta concentration and gamma spectrometry. If statistically significant alpha or beta radioactivity is detected above background ed is verified, that soil sample will- be appropriately analyzed to identify the existing alpha or beta radioisotopes and determine alpha or beta radioisotopic concentration of each identified radioisotope.
8-7
lhe results of the decommissioning contiactor'* hase !! survey will be summarized in a Phase 11 Characterization Report. This report will be submittrd in draf t form to the USAEllA for review. Af ter complMirg their review, the USAfHA will attach the results of their own Phase 11 survey as an ; addendum to the contractor's results and the report will be submitted to the licensee for evaluation. If either evaluation indicates no residual radioactivity above releasable limits, a Termination Radiation Survey Report '
- shall be prepared a, discussed in Section 8.5. If not, 3dditional l decontamination shall be performed and the site characterization process
! repeated for those areas requiring decontamination. l 8.4 INSTRUMENTATION The field instruments to be used during the final survey of the Arill f Reactor sits must be capable of measuring alpha and beta-gamma radioactivity , at the most restricti e release levels specified in Reference 4. Instruments will be * *sted and calibrated in accordance with the specif trations contained : in the irican National Standard, Radiation Protection Instrumentation lest- l and Cali .tton," ANSI N323 1077,- or the most recent revision. j i I Table 8-1 lists a few types of instrunents that could be used during
- final surveys. The detection capabil y of each inv rufrent is also shown in ,
lable 8-1. 8.5 D#.TA DOCUMENTATION Radiation roeasurements and analytical results shall include the following data: e locatict, of the measurement or sample
. Date or dates of incasurements or sample collection e Measured concentration of' the specific nuclides in pCi/g for soi_1 samples 8-8 e Y*'"' '*@'%-"YE'*T*~
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alpha Measurements of radioactivity should be reported as follows: and beta gamma in dpm/100 cm? and gamma at one meter above in 4R/h ll Analytical error at 95% confidence level should be reported for a , analyses or analyst tiamo of surveyor, sampler,
- Analysis date Instrument specifications and calibration dats l
Confidence level, standard error, etc. attached to analytica results
+
Name of person verifying results hi The actual net measured values (including negative valu associated errors shall be reported, Whenever possible, values lower detection (LLD), t he LLD will be provided. 7.4 1 18.5 pCi/g. The than the LLD will be reported in the following manner: following supplemental information shall be included: Description of survey and sampling equipment ts and
+
Survey and sampling procedures, including sampling times, ra e , volumes ' a Analytical procedures '
- Calculation methods
. Calculation of the lower limit of detection
- Calibration procedures 8-10
- - - - __-_-_ _ "~%%~ N
Table 81, Typical survey instruments and detection capabilities f or measuring soil surf ace!.. _t!Rtlidr__ insinetnLOLMeth0L __ DtttC1MJLlutli _ Gross Alpha 1hin-walled, shielded 210' pCi/cmi , very slow scan Pancake G-M 7nS scintillator 2 102 pCi/cm#, source dependent Thin Nat or Caf y 2100 pCl/cm i, source dependent Gross Thin-walled, shielded 500-5,000 pCi/cm?, Beta-gamma pancake G M [ max 2 0.26 MeV 7nS scintillator 300 pCi/mi , source dependent Na! spectrometer 500-5 x 105pCi/cm? lon chamber 2 10' pC i/cm' Phoswich y POD pCi/cm' Intrinsic Germanium 2 200 pCi/cm? portable scintillator 5,000-50,000 pCi/cm 2 s 8-9
Measurements of radioactivity should be reported as follows: alpha and beta gamma in dpm/100 cm' and gamma at one meter above surf ace in pR/h j 1 l Analytical error at 95% confidence level should be teported for all analyses .,
. flame of surveyor, sampler, or analyst ,,
. Analysis date
. Instrument specifications and calibration data Confidence level, standard error, etc. attached to analytical l results l flame of person verifying results The actual net measured values (including negative values) and their associated errors shall be report ed. for values lower than the lower limit of detection (LLD), the tLD will be provided. Whenever possible, values lower than the LLD will be reported in the following manner: 7.4 1 18.5 pCi/g. The following supplemental information shall be included:
Description of survey and sampling equipment Survey and sampling procedures, including sampling iimes, rates, and volumes
. Analytical procedures '
Calculation methods ' l i Calculation of the lower limit of detection I l
. Calibration procedures 8-10
e Discussion of the quality assuratu e pn. gram f or ensuring the quality of results, lhe data shall be presented so that the radiological condition of the f acility and site is completely and accurately depicted and the radiological
' condition can be ascertained without further analysis and manipulation of the data.
s. Based on the results and conclusions of the Phase 1 and 11 Character ization Report s, the deconunissioning cont ractor will prepare a lermination Radiation Survey Report. The report will be reviewed by the USAlllA and provided to the licensee. After the licensee approves the termination Radiation Survey Report, it shall be submitted to the NRC as required by NRC Regulatory Guide 1.86. lhe report will include a description of the survey methods, instrument s, analyses, and an evaluation of the results, lhe t cport is expected to conclude that the AMll Reattor latility and site are suitable for release f or unrestricted use. The Army will make available to the public a plan that will describe the methods by which the reactor building will be dismantled, lhis plan w111 be implemented once the NRC has determined the facility meets unrestricted use criteria. 4 8-11
l REFERENCES FOR ENTIRE DOCUMENT
- 1. Desltiprt Anahsh_RetorLlpr.15 Arn]y_Materiahlrrhtml991_Latterater3 RUfaLLIL2fEler, [GG WM-8979, September 1990.
*' 2. U.S. Nuclear Regulatory Commission, hen 11grlacLivr_19Enllanct.Mjlb Dff DJ!mlM1901ELhraingt iMLSyryryJr11eti_g, NUREG/CR-2002.
- 3. GigtalcirrJIqLion Repjor1_itr_L51_fgmy_,Matet!alLkthunlR9L10b2E0(Or3 Bf3R4ICh E^3 flor, [GG-WM-8978, August 1990.
- 4. U.S. Nuclear Regulatory Commission, Guidelinetfor De(ppiami ail S (Ln_o1 EAC11111ci_anLluit i emsstfrrior_103rleMe f 0LJJure11tkl!d31t_91 krminelimtof usanse_f.or_3yarnqucL..SmitcL_91_SneciaLEucleat Malttial, August 1937.
f 8-12 __ _ __. . _ _ . _ _ _ _ -__ _ _ _-_________ -__________- _____ __ - __ ___ - -}}