ML20151K610
| ML20151K610 | |
| Person / Time | |
|---|---|
| Site: | Waltz Mill |
| Issue date: | 07/25/1997 |
| From: | MORRISON-KNUDSEN CO., INC., PUBLIC SERVICE CO. OF COLORADO, WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20151K593 | List: |
| References | |
| NUDOCS 9708060121 | |
| Download: ML20151K610 (139) | |
Text
.- -. . . . - . - - - - - _ _
l WESTINGHOUSE ELECTRIC l
CORPORATION i
4 l WESTINGHOUSE TEST l
IEACTOR TR-2 I
a
! FINAL DECOMMISSIiONING
'l PLAN
- Revision 0 l July 25,1997
, Prepared By NUMEGA 2
! A Team of:
l Westinghouse Electric Corporation
! Morrison Knudsen l Public Service Company of Colorado f
l 9708060121 970731 DR ADOCK 050000 2
TABLE OF CONTENTS TABLE OF CONTENTS SECTION TITLE PAGE I GENERAL INFORMATION ,
1-1 1.1 License Information.. 1-2 !
1.2 Decommissioning Overview . . .. 1-3 l 1.3 Facility and Site Description. 1-3 l
1.4 Administration of the Decommissioning Plan . . 1-4 References For Section 1. .. 1-5 I 2 DESCRIPTION OF PLANNED REMEDIATION ACTIVITIES. . . 2-1 i 2.1 Decommissioning Method . . . 2-1 2.2 Decommissioning Objective, Activities l Methods and Schedule. . .. 2-2 2.2.1 Decommissioning Objectives. . . 2-2 ,
2.2.2 Decommissioning Activities.. . . 2-2 2.2.2.1 Pre-decommissioning Activities.. . 2-3 !
2.2.2.2 Additional Material Handling Capabilities., . 2-4 i 2.2.2.3 Removal ofHazardous Materials. 2-5 l 2.2.2.4 Reactor Vessel and Biological Shield. . . 2-5 ;
)'
2.2.3 Decommissioning Methods. . . . . . 2-9 l 2.2.3.1 Demolition and Component Removal. 2-10 2.2.3.2 General Surface Decontamination Mehas.. . 2-12 ;
2.2.3.3 Concrete Surface Removal Methods. 2-13 l
4 2.2.3.4 Metal Surface Removal Methods. . 2-14 2.2.4 Decommissioning Schedule.. . . . 2-15 2.3 Decommissioning Work Controls . . 2-16
) 2.4 Decommissioning Organization and Responsibilities.. 2-17 1 2.4.1 Procedures.. . . . 2-18 2
2.5 Contractor Assistance.. . .. . 2-19 2.6 Training Program.. . 2-20 2.6.1 General Site Training. 2-20 4
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REVISION 0 !
' I TABLE OF CONTENTS i
SECTION TITLE PAGE 2.6.2 Radiation Worker Training. . 2-20
- ' 2.6.3 Respiratory Protection Training. .
2-21
! 2.7 Optional Decontamination and . . . 2-22
- Dismantlement Activities within the TR Containment Building 2.7.1 Sub-pile Room . .. . . 2-22 2.7.2 Rabbit Pump Room . . . . 2-23 2.7.3 Test Loop Cubicles.. .
. 2-23 2.7.4 Test Loop Dump Tank Pits.. . . 2-24 2.7.5 Utilities.. . . 2-25 2.7.6 Primary Coolant Pipe Tunnels.. . 2-25 2.7.7 Transfer Canal . . 2-26 2.7.8 Containment . . . 2-27
{
References for Section 2. . . . 2-28 I 3 PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY. 3-1 3.1 Facility Radiological Status.. l
. 3-1 1 3.1.1 Facility Operating History. . . .. 3-1 1 3.1.2 Current Radiological Status of Facility. .. 3-1 3.1.2.1 WTR Structures.. 3-2 l
3.1.2.2 WTR Systems.. . . . . . 3-11 3.2 Radiation Protection Program.. . . . 3-20 i 3.2.1 Radiological Surveillance and Work Area Controls. . 3-21 l 3.2.1.1 Radiological Evaluations.. . . . . . .. 3-21 1 3.2.1.2 Radiation Work Permits.. . . . 3-21 3.2.2 Access Control . .. . 3-22 3.2.3 Facilities and Equipment .. 3-22 3.2.4 Exposure Control . . 3-23 3.2.4.1 External Whole Body Monitoring . .. 3-23 3.2.4.2 Special Monitoring . .. 3-23 3.2.4.3 Skin Monitoring.. .. . 3-24 3.2.4.4 Internal Exposures. . . . .. . 3-24 l
i 11 REVISION 0 l
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TABLE OF CONTENTS l l
SECTION TITLE PAGE 3.2.5 Respiratory Protection Program. . . .. 3-24 3.2.5.1 Respirator User Qualification. 3-25 3.2.5.2 Respiratory Protection Equipment Description.. . . . 3-25 l i and Selection '
- 3.2.5.3 Equipment Inspection and Maintenance. . .. 3-25 ,
3.2.6 Radioactive Materials Control Program . 3-25 i i 3.2.6.1 Radioactive Material Storage. .. .
3-25 '
3.2.6.2 l Contamination Control Program . .
3-26 3.2.6.3 Byproduct Matedal Control. . 3-26 :
3.2.7 Ensuring that Occupational Radiation Exposures are.. . 3-26 ,
As Low As Reasonably Achievable 3.2.7.1 Management Commitment .. . .. . 3-26 3.2.7.2 Radiation Safety Committee.. . .. . 3-27 !
3.2.7.3 Radiological Performance Goals. . . 3-27 3.2.7.4 Plans and Procedures. .. 3-27 3.2.7.5 Radiological Work Planning.. . 3-27 3.3 Radioactive Waste Management. . . .. 3-29 3.3.1 Radioactive Waste Processing.. . . . . 3-30 3.3.1.1 On-Site Radioactive Waste Volume Minimization. . 3-30 3.3.1.2 Off-Site Shipments of Radioactive Materials. . ,, 3-30 for Further Processing 3.3.1.3 Liquid Waste Processing System.. . . .. 3-30 3.3.1.4 Local Ventilation . . . . . 3-31 3.3.2 Radioactive Waste Disposal. . .. . 3-31 !
3.3.2.1 Waste Classification.. . . .. . 3-31 3,3.2.2 Waste Packaging, Transfer and Storage . .. 3-31 3.3.2.3 Waste Transportation.. 3-32 l
3 3.3 Disposal ofNon-Radioactive Waste.. . . . 3-33 ;
3.3.4 Release of Material for Unrestricted Use.. . 3-33 3.3.5 Hazardous Waste.. . .. ... 3-33 3.3.6 Mixed Waste . .. . .. 3-33 3.4 Accident Analysis. . . .. 3-34 3.4.1 Assumptions. . . . 3-34 3.4.2 Dropping of a Contaminated Concrete Block / Rubble. . 3-35 Accident 3.4.3 Fire / Explosion Accident.. . .
3-35 3.4.4 Canal Sediment Criticality and Handling . . . 3-35 3.4.5 Rupture of a HEPA Vacuum Bag . . 3-36 References for Section 3. . . .. . . . . 3-37 :
t REVISION 0
TABLE OF CONTENTS SECTION TITLE PAGE 4 PROPOSED FINAL SURVEY 4-1 l
5 FUNDING.. .
5-1 l References for Section 5. . 5-2 '
l 6 TECHNICAL AND ENVIRONMENTAL. 6-1 SPECIFICATIONS 7 QUALITY ASSURANCE PLAN. . . 7-1 8 ACCESS CONTROL PLAN. . 8-1 8.1 Current Provisions. . 8-1 8.2 Access Control Plan.. 8-2 8.2.1 WTR Access Control Organization. . 8-2 8.2.2 Physical Security Measures. . 8-2 8.2.2.1 Physical Barriers. . 8-2 8.2.2.2 Access Authorization. 8-2 8.2.3 Communications. .. . 8-3 8.2.4 Procedures.. . . .. 8-3 l 8.2.5 Changes to Current Program.. 8-4 8.2.6 Access Control Transition.. . . . 8-4 APPENDIX A Technical and Environmental Specifications for Westinghouse Test Reactor iv i REVISION 0
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TABLE OF CONTENTS TABLE LIST OF TABLES PAGE 2-1 WTR Facilities, Decommissioning Activities and Estimated Worker Exposure . .. .
2-29 2-1(A) List ofTR-2 Miscellaneous System and Components Considered. . . 2-30 i 2-2 Equipment Selection Matrix . .
2-31 3-1 Estimate of Radioactive Waste Volume. .. 3-38 t
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TABLE OF CONTENTS FIGURE LIST OF FIGURES PAGE l-1 Map of Area Surrounding Waltz Mill Site.. . . 1-6 1-2 Overall Map of Waltz Mill Site. . .. 1-7 1-3 WTR Area . .
. . . . . . 1-8 1-4 Cross-Sectional View of WTR Looking East.. . 1-9 2-1 Access Control Points To WTR .. . . . 2-32 2-2 Conceptual Drawing of Rail Transpon . . 2-33 2-3 Conceptual Plan Haul Cart . . 2-34 2-4 Conceptual Drawing One Piece Reactor.. . 2-35 Vessel Removal 2-4 Area Plan For One Piece Removal From.. .. . . 2-36 Containment 2-4 Area Plan For One Piece Removal /.. . .. 2-37 Loading Out To Truck 2-5 Elevations . .. . . .. .. .. . . 2-38 2-6 Upper Portion Removal ofBioshield . . 2-39 By Sectioning 2-7 Mid Portion Removal Of Bioshield Section.. 2-40 2-8 Reactor Vessel Removal Sections.. . . . 2-41 2-9 Removal Of Lower Bioshield Section.. . 2-42 2-10 Removal of Lower Internals . . . . . 2-43 2-11 Lower Reactor Bioshield Section Removal. 2-44 2-12 WTR Decommissioning Schedule. .. . 2-45 vi RET 1SION 0
TABLE OF CONTENTS FIGURE LIST OF FIGURES PAGE 2-13 WTR Decommissioning Project. .. . . 2-46 Responsibility Matrix 2-14 Location of Transfer Canal.. . . 2-47 2-15 Transfer Canal Decontamination.. .. . . 2-48 Sediment Removal Concept 2-16 Transfer Canal Decontamination.. .. . .. 2-49 f
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TABLE OF CONTENTS !1 LIST OF ACRONYMS ALARA As Low.As Reasonably Achievable -
ANSI American National Standards Institute CFR Code of Federal Regulations C.RD Comrol Rod Drive '
CRDM Control Rod Drive Mechanism DOT Department of Transportation l EPA Environmental Protection Agency !
GST General Site Training HEPA High Efficiency Particulate Air HP Health Physics I LSA Low Specific Activity l
MDA Minimum Detectable Activity )
MWt Megawatts Thermal !
NIOSH National Institute of Safety & Occupational Health I NIST National Institute of Standards and Technology i
NRC Nuclear Regulatory Commission l NSD Nuclear Services Division NVLAP National Voluntary Laboratory Accreditation Program ,
PADEP Pennsylvania Depanment of Environmental Protection l PCM Project Control Manual i PMP Project Management Plan PQP Project Quality Plan ]
QA Quality Assurance j RP Radiation Protection !
RSO Radiation Safety Officer RWP Radiation Work Permit RWT Radiation Worker Training SEG Scientific Ecology Group SNM Special Nuclear Material SRD Self-Reading Dosimeters TEDE Total Effective Dose Equivalent TLD Thermoluminescent Dosimeter TR Test Reactor WTR Westinghouse Test Reactor viii REVISION 0
i i GENERAL INFORMATION l l
j SECTION 1 l
GENERAL INFORMATION l
The Westinghouse Electric Corporation (Westinghouse) Test Reactor (WTR) is located on the !
Waltz Mill Site near Madison, Pennsylvania. The retired WTR is currently licensed under
! Nuclear Regulatory Commission (NRC) License TR-2. The balance of the Waltz Mill Site is i
licensed and operated under NRC License SNM-770.
) l 1
Westinghouse has developed a detailed Decommissioning Plan (Plan), based on a Conceptual Decommissioning Plan (Ref.1), to address the activities required to termmate the TR-2 License.
It is considered reasonable and prudent that the activities required for license termmation are:
removal of the remaining reactor vessel internal contents, the reactor vessel, and the biological
, shield. Following these decommissioning activities, Westinghouse will request transfer of the j remaining residual radioactivity and WTR facilities to the SNM-770 License.
This Plan 1
, describes these decommissioning activities and the required interfaces with the SNM-770 licensed
- site.
. Prior to or during TR-2 license termination, the SNM-770 License will be amended to include
- the plans and costs for remediation of the stmetures, materials, and equipment transferred from the TR-2 License, Future use of these structures, materials, and equipment shall be in
, accordance with the SNM-770 license conditions and site procedures controlling occupational exposure and exposure to the public.
This Plan has been prepared using Draft Regulatory Guide DG-1005, " Standard Format and Content for Decommissioning Plans for Nuclear Reactors" (Ref. 2) and the applicable regulatory l requirements associated with 10 CFR 50.82(b). Although DG-1005 is still in draft form, it is considered appropriate for the development and general format of the Plan. The standard fo uat.
of DG-1005 has been slightly altered for consistency with the Waltz Mill Facility SNM-770 Remediation Plan, previously submitted to the NRC on November 27,1996 (Ref. 3).
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M 1-1 REVIS90N 0
1 GENERAL INFORMATION 1
1.1 LICENSE INFORMATION License: TR-2 +
1 Docket Number: 50-22 i I
Location of Use: Westinghouse Electric Corporation I Waltz Mill Site Interstate 70 - Madison Exit 25A I P.O. Box 158 I Madison, PA 15663 $
1 Licensee of Use: Westinghouse Electric Corporation P.O. Box 355 {
I Pittsburgh, PA 15230 i Licensee
Contact:
. Mr. A. Joseph Nardi Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, PA 15230
. The initial WTR operating license was issued on June 19, 1959. Amendment . Number 1 to the operating license, dated January 8,1960, authorized maximum thermal power to. be raised from i
20 MWt to 60 MWt. Westinghouse informed the NRC that the WTR had permanently ceased
{'
operations on March 22,1%2. The following license amendments were issued after permanent cessation of operations.
1 AmeMment Number 2 dated March 25,1%3 - The TR-2 License was amended to allow
_ possession, but not use of the reactor (Possession Only License).
. . 1
- AmeMment Number 3 dated April 22,1970 - The TR-2 License was amended to transfer the Truck Lock Building to the SNM-770 License.
AmeMment Number 4 dated June 24,1970 - The TR-2 License was amended to transfer the Facilities Operations Building to the SNM-770 License.
AmeMment Number 5 dated April 17,1974 - The TR-2 License was amended to extend the license termination to November 30,1993. Since then a timely license renewal letter was sent to
- 11 e NRC on December 8,1992 requesting license extension to November 30, 2003. NRC action
.is pending on this request.
AmeMment Number 6 dated June 14,1993 - The TR-2 License was amended to transfer the three WTR Basins (No.1, 2, and 3) to the SNM-770 License.
l 1-2 REVISION 0 l i
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GENERAL INFORMATION 1.2 DECOMMISSIONING OVERVIEW This Plan describes the objectives, activities, and controls that will apply to the decommissioning of the WTR. The ultimate objective is to terminate the TR-2 License. To accomplish this, the following are the major decommissioning activities:
I
- 1. Remove the remaining reactor vessel internal contents, the reactor vessel, and the biological shield.
. 2. Provide the NRC with sufficient documentation to demonstrate that license termination requirements have been met. This would include all documentation that is required for j.
tansfer of the remaining residual radioactivity and WTR facilities to the SNM-770 License.
4
{ Additionally, decontamination and dismantlement activities of other stmetures and equipment associated with TR-2 may be performed in accordance with this plan. Those activities not !
3 completed under this plan will be completed after being transferred to the SNM-770 license. The
- approved acceptance criteria associated with the retired facilities in the SNM-770 Remediation Plan j will also be used for these other areas.
j 1.3 FACILITY AND SITE DESCRIFflON
- The Waltz Mill site is located approximately 30 miles southeast of Pittsburgh in Westmoreland County, Pennsylvania (see Figure 1-1). The site is approximately 850 acres and is located about i three miles west of the town of New Stanton between the towns of Madison and Yukon (see Figure
- 1-2). The WTR facility is located in the nonhwest ponion of the Waltz Mill site, nonh of the G
- Building (see Figure 1-3). >
2 The Waltz Mill site is operated by the Nuclear Services Division of the Westinghouse Energy
- Systems Business Unit. The WTR is maintained under NRC License Number TR-2 (Possession Only), encompassing the requirements of 10 CFR 50. The WTR license includes the reactor l
, stmeture, reactor systems, the reactor containment building, the rabbit pump room, the sub-pile room, the polar crane, and the WTR ponion of the transfer canal.
The WTR was a low pressure, low temperature, water cooled 60 MWt reactor housed in a l cylindrical vapor containment structure (see Figure 1-4). Since permanent shutdown in 1%2, all l fuel and some of the reactor internal contents have been removed from the reactor vessel and from the Waltz Mill site. The reactor vessel has been drained of all water and the vessel head is secured on the vessel. The Site, including the WTR facility, has been extensively characterized ar.d is controlled to not pose a threat to the health and safety of the site worker or the general public.
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GENERAL INFORMATION 1
1.4 ADMINISTRATION OF THE DECOMMISSIONING PLAN This Deconunissioning Plan provides sufficient detail of the WTR decommissioning activities to i allow NRC review and approval. The provisions of 10 CFR 50.59(e) shall apply to the NRC i j approved Decommissioning Plan and the criteria to be used in evaluating changes to the Plan will ,
be included in project procedures.
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A GENERAL INFORMATION REFERENCES FOR SECTION 1 1
- 1. Westinghouse letter, Nardi to NRC, dated April 7,1997;
Subject:
" Submittal of Remediation Plan for the Westinghouse Test Reactor, USNRC License Number TR-2, Docket 50-22."
i
- 2. Westinghouse letter, Nardi to Bellamy (NRC), dated Nosember 27, 1996;
Subject:
i " Submittal of Remediation Plan for the Westinghouse Waltz Mill Site, USNRC License Number SNM-770, Docket 70-698."
- 3. NRC Draft Regulatory Guide DG-1005, " Standard Format and Content for Decommissioning Plans for Nuclear Reactors," September 1989.
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1-5 REVISION 0 j
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. GENERAL INFORMATION 4
Figure 1-2 OVERALL MAP OF WALTZ MILL SITE 1
1 WTR AREA CENTRAL -EAST SITE AREA ngure 1-3 OPERATIONS AREA a
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GENERALINFORMATION Figure 1-3 WTR AREA i,
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1-8 REVISION 0
GENERAL INFORMATION 1 FIGURE 1-4 i 4
CROSS-SECTIONAL VIEW OF WTR LOOKING EAST '
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1 i
l 1-9 REVISION 0
CH01CE OFDECOMMISSIONINGMETHOD AND DESCRIPTION OFA CTIVITIES l 3
i SECTION 2 i CHOICE OF DECOMMISSIONING METHOD AND DESCRIFFION OF ACTIVITIES 2.1 DECOMMISSIONING METHOD l
- Decommissioning, as described in this Plan, will be accomplished by removal and disposal of the i remaining reactor vessel internal contents, the reactor vessel, and the biological shield. The balance of the WTR facility components and the remaining residual radioactivity will be transferred to the !
SNM-770 License. There are no radiological limits applicable to the transfer of structures,
)
- materials, and equipment to the SNM-770 License, other than the radioactive materials possession limits specified in the SNM-770 License. Prior to the transfer, the SNM-770 License will be amended as necessary to include the remaining WTR associated radioactive material inventory.
Additionally, any other document revisions required as a result of this transfer will be perfonned.
Future use of these structures, materials, and equipment shall be appropriately maintained in accordance with the SNM-770 license conditions and site procedures controlling occupational and i pubhc exposure.
l In addition to removing the reactor vessel internal contents, the reactor vessel, and the biological shield, decontamination and dismantlement activities may be performed on other structures and
- equipment located within the WTR containment building. These other activities are not required for WTR deconunissioning; however, they are addressed herein as optional activities that may be j undertaken under the authority of the TR-2 Decommissioning Plan, prior to transfer of remaining residual radioactivity and WTR facilities to the SNM-770 License. The approved acceptance criteria associated with the retired facilities in the SNM-770 Remediation Plan will also be used for these other areas.
4 i Precedent for transferring the residual radioactivity to the SNM-770 License has already been I established by Amendment Numbers 3, 4, and 6 to the TR-2 License. These Amendments l
. transferred previous WTR facilities to the SNM-770 License (Truck lock Building, Facilities l Operations Building, and WTR Basins).
2-1 REVISION 0
CIIDICE OF DECOMMISSIONINGMETHOD AND DESCRIPTION OFACTIVITIES 2.2 DECOMMISSIONING OBJECTIVE, ACTIVITIES, METHODS AND SCHEDULE 2.2.1 Decommissionine Obiectives The objective of this Decommissioning Plan is to outline the activities for removal of the WTR reactor vessel internal contents, the reactor vessel, and the biological shield, to the point where the TR-2 License can be terminated by transferring the remaining residual radioactivity and WTR i facilities to the SNM-770 License. Decommissioning will be by removal, dismantlement, '
decontamination, release of clean items and disposal of contaminated waste 2.2.2 Decommissionine Activities The general activities needed to complete the Plan objectives are:
Remove the remaining reactor vessel internal contents, the reactor vessel, and the biological shield.
Prepare the decommissioning generated material for release or disposal; either decontamuute and release as non-radioactive waste, or package for transport as radioactive waste.
Ship all radioactive waste off-site to a licensed waste processor or disposal facility. In the event that no acceptable licensed disposal facility is available, waste may be retained onsite or, after processing, returned to the site for interim storage. )
Determine the residual radioactivity remaining and prepare the necessary amendments to the SNM-770 License.
Request transfer of the remaining residual radioactivity and WTR facilities to the SNM-770 i License.
Request termination of the TR-2 License.
The Plan includes examples of decontamination techniques, equipment and materials which may be used, a schedule, special training requirements for workers, radiation protection and occupational safety and health pmetices. Selection of decommissioning methods is heavily influenced by worker and public ALARA considerations. A list of WTR facilities, planned decommissioning and i decontamination activities and estimated worker exposure (person-rem) is presented in Table 2-1.
Work plans will be prepared to address issues such as asbestos, lead, or other known hazardous materials in the area of work. The final decommissioning methods will utilize the best, most economical means to minimize hazardous, mixed and radioactive waste volume requiring licensed disposal. From the standpoint of cost-effectiveness, contaminated equipment, materials, etc., may be decontaminated, allowing release for unrestricted use, or packaged for transport and disposal.
This Plan allows flexibility in the choice of decontamination procedure / technique and sequence.
2-2 REVISION 0
CIIOICE OF DECOSfMISSIONING hfETilOD AND DESCRIPTION OFA CTIVITIES 2.2.2.1 Pre-decommissionine Activities Two alternative metho<is for removing the WTR reactor vessel are under consideration that potentially affect pre-decommissioning activities. These two options are one-piece removal and multiple piece removal. The one-piece removal option involves cutting an opening in the containment building and lifting the reactor vessel and pan of the biological shield out of the containment building with an external crane. The multiple piece removal option involves sectioning the reactor vessel and the biological shield concrete into pieces that can be removed through the tmck lock with the existing overhead crane. As discussed below, these alternatives have different impacts on activities to upgrade the existing crane, maintain integrity of the containment building, and install a filtered ventilation system.
Access Control Initial access to the WTR facility will be established through the existing air locks that separate the WTR from the G Building Annex. This access could be used for the installation of a HEPA filter system to the existing containment building and for any required repairs to the interior truck lock door. Following HEPA installation and operational verification of the filtration system, the majority of equipment and material access to the WTR will be through the adjacent tmck lock on the north side of the reactor, except for any materials removed through a temporary containment building !
access opening, if the one-piece removal method is used. I The east air lock will continue to be used as the main control point for personnel access to the containment building. A change area will be provided at the entrance to the east air lock in the annex to route personnel upstairs and out through the annex building (see Figure 2-1). Personnel access to the containment building may also be provided through the imck lock.
HEPA Filtration / Ventilation System A HEPA filtration / ventilation system will be installed. This system will be capable of creating a negative air pressure within the containment building when personnel access airlock doors are open.
In addition, this system will be capable of maintaining an inward airflow within the containment building during times when a large component removal hole (if installed) is open in the containment building.
Tmck Lock Door Electrical service will be re-established and repairs made to the truck lock door motor and hoist to allow controlled equipment and material access to and from the containment area.
Temporary Utilities Temponry lighting and power will be installed in accordance with applicable requirements, as well as local safety codes. Some existing electrical systems may be used, after inspection and repairs.
Polar / Mobile Crane The polar crane and components may be repaired and/or upgraded, as necessary, to allow for safe operation throughout the decommissioning activities. Prior to use, the manufacturer or qualified 2-3 REVISION 0
C110 ICE OF DECOMMISSIONING MET 110D AND DESCRIPTION OF ACTIVITIES inspector will cenify the crane and components for safe operation, including performance of necessary load tests.
An alternative to using the overhead crane is using a mobile crane operated from outside of the containment building for reactor vessel one-piece removal. This requires that a hole be cut in the containment building roof to allow the crane to access the sectioned components for lifting. If cutting is required after the containment vessel is breached, additional engineering or administrative controls will be used.
Decommissionine Activity and Associated Person-rem Each decommissioning activity has an estimated worker exposure calculated for that task which is dependent on labor loading, decommissioning method, and known radiological conditions. The decommissioning methods selected strive for ALARA exposures to the workers. These estimated doses are presented in Tabic 2-1 at the end of this section. I i
2.2.2.2 Additional Material Handline Capabilities l
l General !
To facilitate safe and efficient material handling capabilities, temporary suppon stmetures may be assembled and installed. This may include providing a method for easy transportation of heavy and/or bulky materials and equipment out of the containment building, as well as providing an additional temporary containment (auxiiiary area) adjacent to the truck lock.
Temporary Transponation System I Heavy and/or bulky materials which requim removal from the containment building, such as !
sectioned concrete, reactor vessel and components, etc., may require additional transponation !
capabilities. The addition of a rail can or similar capacity transportation device in the tmek lock area will allow the safe and efficient removal of the material to a staging area in preparation for transport or to other areas for funher processing. This temporary transpon system will not be required if the reactor vessel is mmoved in one piece. An example of such a transpon system, in this case a rail can, is shown in Figures 2-2 and 2-3.
i Adiacent Auxiliary Area A temporary auxiliary area adjacent to the tmck lock building (shown as Tented Area in Figure 2-
- 2) may be utilized to process material removed from the containment building. This auxiliary area will be covered by a temporary building. The area may be used to decontaminate material, survey and/or sample material, section or segregate clean from contaminated material, and/or package material for transportation to an off site processing location or a licensed disposal facility. This area may be necessary due to space constraints within the containment building and will allow dismantling activities to progress with mimmal interruption.
The temporary auxiliary area will be fully contained and provided with a HEPA ventilation system sufficient to maintain a negative pressure within the area while materials are processed. Procedures 2-4 IEVISION 0
l I
l CH01CE OF DECOSDHSSIONING METIl0D AND DESCRIPTION OFA CTIVITIES and/or work plans will describe acceptable methods for movement of materials into and out of the auxiliary area.
I 2.2.2.3 Removal of Hazardous Materials i Irad Approximately 266,000 pounds (385 cubic feet) of lead in the form of brick, sheet, shot and other casting remain in the reactor area. Some of the lead material may require decontamination prior to '
final disposition. Lead will be surveyed and/or sampled for radioactive contamination in order to segregate clean material from contaminated material. The material will be packaged in transpon ;
containers, as necessary, and removed from the containment building. Contaminated lead may be decontaminated on site or transported to a licensed facility for treatment. Options for the beneficial re-use of lead will be evaluated and the most cost effective method for fmal disposition pursued. !
Lead-Containine Coatines Demolition work performed during the TR-2 decommissioning project may require the removal of {
lead-containing coatings, or remediation in areas where lead dust may have accumulated. Upon identification of these areas, a qualified lead abatement subcontractor, or qualified remediation team
)'
workers, will be used to remove the lead containing coatings or dust.
Any work performed that requires a torch to metal that has a lead-containing coating, will have the coating mmoved by a qualified abatement subcontractor or qualified remediation team worker. The ,
activities will be performed prior to any torch to metal work or grinding of lead containing coatings and will comply with the Waltz Mill Remediation Project Site Specific Safety and Health Manual.
Asbestos Abatement Asbestos containing materials will be removed and packaged for disposal prior to any decommissioning activities in areas where these materials exist, provided these activities can be conducted safely and radiation exposure can be maintained ALARA. Asbestos has been identified ;
in the floor tiles on the operating floor (elevation 16'0"), on several of the intermediate reactor l platforms (elevation 32'3" and 36'71/2"), and in the test reactor piping systems insulation.
Removal and disposal of asbestos will be accomplished by a licensed asbestos abatement contractor. 1 Additional asbestos materials discovered in the course of decontamination activities will be abated by the asbestos contractor, as needed, j I
2.2.2.4 Reactor Vessel and Biolocical Shield General The stainless steel reactor vessel is centered and encased within the biological shield. The vessel is approximately 8 feet in diameter and 32 feet in length, extending vertically from approximately I I
elevation 62' down to the top of the sub-pile room at elevation 29'. Access to the reactor vessel is available from the top, at the reactor head, and the bottom at the vessel's bottom flange. The l interior of the vessel may be accessed by removing either the reactor head or the bottom flange in the sub-pile room. Radioactive contamination is present as surface contamination and component l
2-5 REVISION 0 i
l
CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OF A CTIVITIES
- activation within the reactor vessel. The most feasible means of access will be through the reactor head.
1 Removal of the WTR reactor vessel and biological shield will proceed following either one of two 1 options:
e i
Option 1- One-Piece Reactor Vessel Removal Option 2- Multiple Piece Reactor Vessel Removal l
l' Both of the options are presented in this Decommissioning Plan to allow overall project flexibility.
The final course of action will be determined based on engineering, licensing, and ALARA
- considerations. l l
- Option 1- One-Piece Reactor Vessel Removal Option 1 involves removing the majority of the biological shield, and lifting the entire reactor vessel and internal components intact out of an opening cut into the top of the containment building.
Details of the rigging and lifting are provided in Section 2.2.3.1; a conceptual drawing is provided as Figure 2-4 sheet 1 of 3,2 of 3, and 3 of 3. i Option 1 requires the following actions for one-piece removal of the reactor vessel and internal components:
i a) Remove excess portions of the biological shield; j b) Inject low density grout into the reactor vessel;
.. c) Fix external contamination and prepare the vessel for rigging; d) Cut an opening in the dome of the containment building;
]i e) Lift the reactor vessel and remaining biological shield out of the containment building;
- f) Prepare and ship the vessel to a licensed disposal facility. ;
4 Remove Excess Biological Shield The excess biological shield will be cut from the vessel, removed, arxl staged for final disposition in !
a safe and secure manner. With the excess biological shield removed, the reactor vessel and remaining biological shield will be approximately 32 feet tall by 10 feet square, and will weight I approximately 148 tons.
Iniect Low Density Grout into the Vessel A low density cellular grout (approximately 20-25 pounds per cubic feet wet density) may be used for stabilizing components and fixing contamination inside the reactor vessel. The reactor piping may also be removed and the control rod drive mechanisms will be removed from the reactor vessel. Covers will be positioned and welded to the cut / prepared reactor vessel openings. Once the major openings are sealed, the reactor vessel may be filled with low density grout. Some opening will have to be used to inject the grout.
2-6 REVISION 0
CHOICE OF DECOMMISSIONINGMETHOD AND DESCRIPTION OFACTIVITIES The grout mix, equipment, materials, personnel and methods to be employed for this operation will be substantially the same as those previously used for other large nuclear steam supply system component removals. The grouting equipment will be kept outside the containment as much as possible to avoid contamination and minimize waste volumes.
EiiL External Contamination and Prepare for Rieeine A paint or similar coating will be applied to the outside surface of the remaining biological shield to fix ccntamination in place. This paint / coating will be a high solids encapsulating paint / coating.
(This application has been used in similar processes for steam generator component removal.) The paint / coating may be applied to the surfaces with mimmal surface preparation. In addition to or as a replacement for the painting / coating, the remaining biological shield may be placed in a container or sleeve to control the spread of contamination.
Cut an Opening in Dome of the Containment Buildine A layout plan will be prepared for accurate alignment of the dome cutting operations to minimize the size of the opening required for the large component removal. After the cut layout is marked on the dome, the cut may be made, using torch or equivalent methcd. A temporary closure will then be installed over the opening after the cut is complete; the opening will be uncovered only during actual rigging and lifting of the components. This temporary closure will allow a negative pressure to be maintained in the containment building when the enclosure is installed.
Riccine and Lifting the Reactor Vessel and Remaining Biological Shield After the vessel has been prepared, the outside crane will be positioned and the rigging attached to the reactor vessel and remaining biological shield. The lifting rig will be designed to lift the total calculated loads. Once the lifting arrangement has been attached, the rigging slack will be taken up, and the load transferred to the outside crane. The reactor vessel will then be lifted from the containment building and staged in a safe and secure manner.
l Prepare and Ship the Reactor Vessel / Remaining Biolocical Shield to a Disposal Facility I After the reactor vessel and remaining biological shield have been lifted out of the containment building, it will be prepared for shipping. Either the vessel / biological shield will be modified so i that it becomes the waste package, or it will be placed inside a cask / container. Packaging, shipping, and transportation will comply with all applicable licensing and shipping regulations.
Safety analyses and radiological surveys will be performed, and special permits will be obtained before shipping the vessel / biological shield to a licensed disposal facility, as required.
Option 2 - Multiple-Piece Reactor Vessel Removal j The multiple-piece option involves cutting the biological shield off of the reactor vessel using a diamond wire saw, removing the upper and lower reactor internals, and cutting the upper, middle. l and lower vessel into sections. All of the sections will be within the capacity of the interior polar crane to allow moving the sections from the work area onto a tansport system and then out of the containment building. The process for the multiple piece removal is described as follows:
2-7 REVISION 0
CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OF A CTIVITIES Upper Vessel Internal Components Prior to. removal activities an interim HEPA filtration system, capable of creating a negative pressure within the vessel, will be installed at one or more inspection pons at the reactor head (elevation 61' 81/2", see Figure 2-5). From the sub-pile room (elevation 29' 8"), access ports will be removed from the bottom flange of the vessel to allow installation of HEPA filtration ducting at the bottom of the reactor vessel. The interim HEPA system, at the reactor head, can be removed once ventilation through the bottom flange is established. The reactor head can then be removed using a crane.
The head can be placed on the head stand located on the second platform (elevation 51'0").
Depending on radiological conditions, construction of a temporary containment and air lock over the vessel may be required at the reactor head while the internal components are removed.
Removal of upper internal components from the reactor vessel can be done manually using long handled tools, as appropriate, to maintain exposure ALARA. Figure 2-5 depicts the reactor with internal components in place. Control rods, guides, flanges and piping penetrations will be dismantled with hand tools and/or cutting, as appropriate. Once the reactor internal components are removed, reducing the dose rates within the vessel, access can be allowed provided that exposure can be maintained ALARA. Welded components within the reactor vessel will be removed using appropriate cutting equipment (e.g., plasma torch). A lay down area for the internal components will be located on the platform adjacent to the reactor head (elevation 61'8 1/2"). Debris will be handled manually, again using long handled tools, as appropriate, or lifted out using a crane. Waste containers will be positioned on the platform for material packaging and removal. Filled waste containers will be removed from the platform using a crane and positioned on the transport system for transfer to the staging area.
Unner Platforms and Biological Shield The upper platforms and the upper biological shield will be removed after the upper vessel internal components are removed. Figure 2-6 illustrates how the upper ponion of the biological shield will be removed by sectioning. The sectioning plan is based on the results of concrete core samples, which allows for separation of activated from non-activated concrete. Concrete blocks will be sectioned to stay within the load limits of the crane. Biceks of removed concrete will be moved by the crane and placed on the floor at elevation 16'0" in a designated low background area. The blocks will be surveyed and sampled for contamination and prepared for removal from the containment building. Contaminated blocks will be transponed to an auxiliary area for decontamination or packaging for disposal. Concrete blocks meeting the unrestricted release criteria may be transponed to an appropriately permitted landfill. This procedure will be repeated throughout the removal of the remainder of the biological shield and upper platforms.
Mid Biological Shield Area The mid biological shield area will be removed from the perimeter of the vessel leaving a center square column of concrete around the core of the reactor. This column of concrete will remain, acting as shielding, until removal of the vessel's internal components is complete. The mid 2-8 REVISION 0
CHolCE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES biological shield will be removed from elevation 51'0" down to approximately 34'0" (see Figure 2- ,
7). Some of these sections of concrete blocks will require additional sectioning to remove l
l contamination on the sides nearest the reactor vessel. This activity will take place in the Auxiliary 1 i Area after the blocks are removed from the containment building.
- i The center column containing the vessel will be reduced as shown in Figure 2-8. These blocks of concrete and the portions of the vessel contained within are contaminated, or contain activated
- materials. These sections are not economical to decontaminate or funher volume reduce. Some of ;
i these sections will require special containers to shield higher levels of radioactivity. The l l containerization of this waste will take place inside the containment building. These containers will I l then be transferred out of the containment building by the transport system arxl moved to a j temporary storage area on site or to a licensed disposal site.
I I_nwer Internal Components 4
The lower intemal components and the remainder of the mid biological shield will be removed as I 4
shown in Figures 2 9 and 2-10. Decontamination of a majority of the contaminated / activated j intemal components using existing technology is not feasible and they will therefore be '
j containerized prior to leaving the containment building. The imck lock platform may need to be 1 removed if the lower biological shield and sub-pile area are removed (see the following section). If l it is necessary to remove the lower biological shield and sub-pile sections and, consequently, the j tmck lock platform, a new stmetural steel replacement platform may be required. If the biological shield below elevation 32'3" can be decontaminated without disassembly, the truck lock platform will remain.
i l Iower Biolocical Shield Due to the levels of contamination in areas within the lower section of the biological shield, it may be necessary to remove the entire base as opposed to ponions, or decontaminate in place. This is a decision that will require further consideration as the area is exposed during the decommissioning effort. Figure 2-11 illustrates the methods of removal of this section. The blocks of concrete will be staged and removed as previously discussed with the exception of the utilization of the newly constmeted stmetural steel platform in the place of the removed tmck lock platform. This approach will leave the lower level base elevation at approximately 19'. It will then be detennin d whether further reduction will be necessary. The remaining contaminated portions could be either cut away or decot > ainated in place.
2.2.3 Decommissionine Methods WTR Decommissioning involves removal and disposal of the reactor vessel intemal contents, the reactor vessel, and the biological shield. This includes the following activities:
- 1. Remove and dispose of material as radioactive waste
- 2. Remove, decontaminate as necessary, and release material for unrestricted use (this will generally involve disposal at a landftll or processing at a scrap / recycling facility) 1 2-9 REVISION 0
CHOICE OF DECOMMISSIONINGMETHOD AND DESCRIPTION OFACTIVITIES
. Activities that may be undenaken to dismantle and decontaminate other areas within the
- containment building are described in Section 2.7, and will involve additional decontamination and removal processes. These areas will be left in place and transferred to the SNM-770 License. i Each major equipment item and area will be evaluated to determine the best method (s) for removal, for decontamination, and to determine whether to decontaminate or dispose of as radioactive waste. l Criteria to be used in the evaluations include: availability of a burial facility; the cost of i
- decontamination versus the cost of burial; radiological and occupational hazards involved; and site operations in progress or planned.
i Removal of structures, equipment and components can be achieved using proven i mechanical / thermal cutting and demolition equipment. Mechanical methods such as diamond wire 1
] cutting, saw cutting, concrete scabbling, expandable grout, the use ofjackhammers, and machining !
may be utilized. Thermal methods such as metal cutting with an oxy-acetylene torch method may also be used.
J l
! 2.2.3.1 Demolition and Component Removal i 1 i
} Decommissioning of TR-2 involves removal of the reactor vessel, the biological shield, and the vessel internals.
Methods used for the removal of concrete include jackhammers, expandable grout, concrete saws, j and diamond wire saws. These methods am described as follows: I l
Jackhammer
. Equipment can range in size from hand held units to large hoe rams mounted on tracked excavators. The concrete is degraded through constant pneumatic impact of a chisel pointed bit.
This method works well but it is noisy and produces large quantities of dust and debris. Typically a containment tent is constructed over the area and supplemental roughing filter and HEPA filter ventilation is used to control airborne dust and radioactivity.
Hoe Ram Where large areas of concrete require removal, it may be cost effective to decontaminate the i concrete to acceptable levels by other means and then use large hoe rams to remove the cancrete. !
Appropriate controls will be used to minimize the spread of airborne dust and mdioactivity.
Expandable Grout Expandable grout may be used for demolition of concrete structures or removal of predetermined i layers of concrete from structures. Holes are systematically drilled into the concrete in preparation for the addition of the grout. The grout is then mixed with the appropriate quantity of water and poured into the pre-drilled holes. As the grout hardens, it expands and cracks the concrete apan. l 2-10 REVISION 0
CIIDICE OF DECOMMISSIONINGMETIIOD AND DESCRIPTION OFACTIVITIES Concrete Saws Concrete saws may be used for accurate cutting of concrete for general demolition and dismantlement. Also this meth, d may be used to cut large slabs of concrete for waste volume reduction or packaging.
Diamond Wire Cuttine Diamond wire cutting techniques can be used to remove large segments of concrete. A diamond-j studded cable is circulated by a hydraulic pulley drive system through the concrete, cutting through '
concrete, steel rebar and other steel members in the concrete. Hydraulic cylinders control the l
tcnsion of the cable. Holes are drilled through the concrete to enable stringing the cable into cutting l target areas that would otherwise be inaccessible. Water applied to cool and lubricate the cable also aids in control of airborne dust. A slurry collection system is installed to collect contammated
{
I cutting slurry, decant the slurry and recycle the water.
Pipe Removal Various reactor system pipes and sample loop piping will be removed as part of TR-2 decommissioning. Steel pipes are generally removed using mechanical or thermal cutting methods, such as hand-held band or reciprocating saws and oxy-acetylene cutting torch. Commercially available oxy-lance and plasma are cutting methods may also be used. Plasma are cutting equipment can be track-mounted and operated remotely, minimizing personnel exposure in high )
radiation areas. It also can cut underwater. 1 Riccine and Liftine Plans will be developed for removing equipment and material from inside the containment building to a safe and secured area outside of the containment building. These plans will include the integration of equipment, methodology, and training of personnel to enhance total safety as much as practical. All rigging and lifting will be performed in accordance with industry standard safe practices and lifting equipment will be designed to comply with ANSI /ASME specifications.
The rigging and lifting method selected will depend upon whether the one piece, or the multiple piece reactor vessel removal approach is selected. The one piece removal method involves a large 4
capacity external crane to lift the reactor vessel and large slabs of the biological shield through an opening cut in the containment building dome. The advantages of the one piece removal are that less cutting and packaging is required and worker exposures are reduced. However, this method involves greater rigging challenges and a hole has to be cut into the containment building. The multiple piece removal method involves cutting the reactor vessel and biological shield into pieces small enough to be handled by the existing interior polar crane. The advantages are lower waste volumes and ease of handling / packaging smaller pieces with the existing polar crane, and then moving them to the truck lock and out for packaging and shipment offsite.
Radiological Control /Eauipment Decontamination Equipment will be checked for residual contamination before exiting designated restricted areas. Any equipment utilized within a designated restricted area will be decontaminated 2-11 REVISION 0
CII0lCE OFDECOMMISSIONINGMETIIOD AND DESCRIPTION OFACTIVITIES before removal from the work area. The restricted area will be demarcated by flagging, physical barricades or fencing as deemed appropriate.
Loadine/Shippine Loading of shipping containers and hauling equipment will be controlled to minimize contamination on external surfaces. Containers / loads will be secured / covered. Material designated for off-site disposal will be placed in packagings which meet DOT requirements, and staged in a secured area to prevent inadvertent removal from the site.
2.2.3.2 General Surface Decontamination Methods The methods described below are typi2', other processes and technologies may be used.
Strippable Coatines I Strippable coatings may be used to assist in the removal of loose radionuclides from large l surface areas. Strippable coating is a simple, effective means of removing loose radionuclides or protecting areas that may possibly be contaminated during scheduled work activities. Once the surface is dry, the strippable coating serves as a barrier preventing radionuclides from reaching the surface below. If the barrier becomes contaminated, it can be stripped away, or the radionuclides can be sealed in place with a second layer and subsequently stripped away.
Any method normally used to apply coatings (airless sprayers, paint ro!!ers or bmshes) may be l used to apply strippable coatings. l Vacuumine/Scrubbine/Winine l These techniques are generally used when gross loose radionuclides are visible on the targeted surface (s). Vacuum operations use systems equipped with a HEPA filter. If a wet vacuum is required for liquid retrieval, the vacuum system also includes an automatic water shut-off l system to prevent destruction of the HEPA filter when the unit is full. Scrubbing and wiping techniques are used where access is limited or can not be reached with a vacuum unit. It should be noted that vacuuming can be used on any type of surface but sembbing/ wiping are normally used on smooth, non-porous surfaces.
Pressurized Water Pressurized water spraying may be used for general area decontamination or decontamination of items and components in a confined spaca. This method will only be utilized in areas where the spent water can be directed into a drain, sump or some other means of collection. The contaminated water will be treated and monitored to ensure compliance with discharge limits before discharge. Descriptions of pressure spray methods follow:
Low Pressure Spray (Power Wash)
Water is sprayed on the surface to be decontaminated with the objective of removing loosely adhered contamination. This technique is effective on coated surfaces that allow the contamination to be removed easily. Water pressure is generally in the 1,500 to 5,000 psi range with water consumption typically 3 to 6 gallons per minute.
2-12 REVISION 0
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CHOICE OFDECOMMISSIONINGMETIIOD AND DESCRIPTION OFACTIVITIES
. High Pressure Spray (Hydrolaser)
A powerful stream of water is applied to the surface in a side to side, top to bottom fashion. This method is used on large surfaces and complex structures or equipment. +
This technique is effective on coated and uncoated contaminated surfaces. The water i
can be applied in various temperature ranges and the addition of chemical agents may
, increase the overall decontamination factor; chemical agents will be carefully selected to reduce the possibility of creating mixed wastes. High pressure sprays typically
- operate over a pressure range of 5,000 to 20,000 psi, with a water consumption rate of
- about 5 gallons per minute.
, Ultra High Pressure Ultra High Pressure water can be utilized in two ways. The first method is to direct a q precise stream of water from a multi-jet rotating nozzle at the target surface. This j method is capable of removing loose as well as fixed contamination from the surface.
The second method utilizes a single nozzle that is capable of cutting material from the i
surface. Water consumption varies with nozzle selection; rates from 2 to 6 gallons per minute are typical with water pressure approaching 40,000 psi.
Steam Lance Saturated steam is directed on the surface to be decontaminated. Crystalline materials ,
can be solubilized and particulates removed using this technique. This method may be '
useful when the surface is sludge or oil coated.
2.2.3.3 Concrete Surface Removal Methods Scabbline When coating removal and/or surface removal is required, scabbling may be the preferred decontamination method. Scabblers remove the surface by impacting the area with air driven tungsten carbide tipped bits. Scabbler sizes range from single-piston units suited for small constricted or isolated areas, to multi-piston units designed for operation in large open areas.
Surface removal can vary from a light ; ingle pass removing 1/16 inch to multiple activities removing 1 inch or more. HEPA filter vacuum units will be attached to shrouds around the scabbler heads tc, control airborne radioactivity, where necessary.
Scarification Scarification is the process of removing a surface layer of material from concrete floor slabs or similar surfaces. This equipment is generally utilized for projects where wide open floor areas are contaminated and require surface removal. A scarifier is a mechanically powered (electric, gas or propane) device that removes surface layers of material with a rotating drum equipped whh tungsten carbide tipped cutters. When the unit is operated the bits are forced against the surface at a predetermined depth and lateral speed.
2-13 REVISION 0
CHOICE OF DECOMMISSIONING AIETHOD AND DESCRIPTION OFA CTIVITIES A HEPA filtered vacuum system operated in conjunction with a vacuum shroud attached to the scarifier is used in controlling airborne radioactivity during operation. Typical surface removal depths vary between 1/16 to 1/4 of an inch per pass.
Grindine On a smaller scale, hand held grinders can also be used to remove surface coatings or concrete.
Needle Gun A needle gun operates by pneumatically driving specially hardened needles into the surface being cleaned. The needle gun is designed to remove surface material from small areas or restricted spaces. The process takes place within a vacuum shroud, preventing the escape of dust, debris and airborne contamination. The vacuum shroud is connected to a HEPA filtered vacuum system that provides the negative pressure required.
Abrasive Blastine Abrasive blasting is a preferred metal surface removal method, and is described below.
However, it can also be effectively used for concrete surface removal. Blastrac (distussed in following section) is commonly used for decontamination of concrete floors.
2.2.3.4 Metal Surface Removal Methods Abrasive Blast Coating and/or surface removal can be achieved using an abrasive blast method. This technique is capable of removing loose and fixed contamination with a high production rate.
Abrasive blast techniques use non-hazardous abrasive material suspended in a medium (air or water) that is propelled against the targeted surface. The result is a fairly uniform removal of
'mface material. High production rates are common. Overhead and vertical surfaces can be decontaminated with relative ease. Depending on the equipment used and radionuclide levels encountered, the blasting medium may be reused.
Blasting media include sand, steel, aluminum oxide, walnut shells and plastic. Supplemental HEPA filter ventilation is used when necessary to control airborne dust and radioactivity. '
Recycled Abrasive Blast (Blastrac)
Shot blasting is an airless method that strips, cleans, and prepares the surface for coating application. Surface removal can be achieved by selecting the proper shot size and residence time. The shot is propelled at the surface using a centrifugal blast wheel. As the wheel spins, the abrasive is hurled from the blades, blasting the surface with a barrage of media. The Lt,rasive is continuously recycled using a vacuum system in conjunction with a separation system.
Supplemental HEPA filter ventilation is required to control airborne dust and radioactivity.
2-14 REVISION 0
CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES 2.2.4 Decommissioning Schedule
= The WTR Decommissioning Project is currently scheduled from February 1998 to 2003. The decommissioning project schedule assumes NRC approval of the Decommissioning Plan by January 1998. See Figure 2-12, entitled "WTR Decommissioning Schedule."
Changes to the schedule may be made at Westinghouse's discretion as a result of resource allocation, availability of a radioactive waste burial site, interference with ongoing Waltz Mill operations, ALARA considerations, further characterization measurements and/or temporary on-site radioactive waste storage operations.
2-15 REVISION 0
1 CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES :
2.3 DECOMMISSIONING WORK CONTROLS l
Work controls will be established to ensure remediation work is safely performed in accordance with the Decommissioning Plan, Waltz Mill license requirements and established procedures.
4 A Project Management Plan (PMP) will be prepared that describes the approach and methods to be used to ensure the successful decommissioning of Waltz Mill facilities. The PMP will provide descriptions of the management philosophy, approach, and techniques to be used on the project. The system of work controls described above will be proceduralized in a Project
' Control Manual (PCM), which will include implementing procedures and supporting l information for preparation of the Work Breakdown Structure, Work Specifications and Work 1 Packages, in accordance with requirements of the Decommissioning Plan.
I A General Work Specification will be developed to establish the basic requirements and .
provide the planning information for the performance of work activities. In addition to the l
General Work Specification, other Work Specifications may be prepared for activities that require special controls (e.g., water treatment).
Work Packages will be prepared based upon the Work Specifications and will contain the detailed instructions for accomplishing the dermed tasks.
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2-16 REVISION 0
CII0lCE OFDECOMMISSIONINGMETIIOD AND DESCRIPTION OFACTIVITIES 1
2.4 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES The Decommissioning organization is integrated into the existing Westinghouse Waltz Mill
- facility organization and complies with the existing license and applicable regulatory requirements.
The direct responsibility for operational oversight of activities conducted under the TR-2 License and the Waltz Mill Site Radiation Protection Program rests with the Waltz Mill Site Manager (current title is Manager, Resources and Support Operations) who reports directly to i
the Division Gen:ral Manager (current title is NSD General Manager). The Waltz Mill Site Manager will continue to have overall responsibility for the facility and the functional groups for: operations, engineering, industrial hygiene, safety, security, environmental compliance, j facilities support, and radiation protection.
Reporting to the Waltz Mill Site Manager is the Radiation Protection Manager (current title is i
Industrial Hygiene, Safety and Environmental Compliance Manager) to whom the Radiation Safety Officer (RSO) reports. The RSO is responsible for the establishment and guidance of programs in radiation protection. The RSO also evaluates potential and/or actual radiation exposures, establishes appropriate control measures, approves written procedures, and ensures compliance with pertinent policies and regulations. Under the RSO's direction, health physics personnel administer the established site policy, collect samples, perform analyses, take measurements, maintain records, and generally assist in performing the technical aspects of the radiation protection program. The health physics staff reports directly to the RSO. The RSO will be supported by adequate staff, facilities and equipment and will hold a position within the organizational stmeture providing direct access to senior management.
The Remediation Team Program Manager repons to the Waltz Mill Site Manager. The Remediation Team Program Manager will coordinate the elements of the functional groups of the Waltz Mill decommissioning organization, Remediation Team, and deconunissioning contractors, as it applies to decommissioning activities. The Remediation Team reports to the Remediation Team Program Manager.
The existing Radiation Safety Committee required under the SNM-770 License will monitor decommissioning operations to ensure they are being performed safely and according to federal, state, and local regulatory requirements (NRC, EPA, PADEP, DOT, etc.). Members of this committee are appointed by the Division General Manager. The Radiation Safety Committee will review major decommissioning activities dealing with radioactive material and
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radiolegical controls. In addition, the Radiation Safety Committee will review and approve changes to the Decommissioning Plan that do not require prior NRC approval.
4 2-17 REVISION 0
CHOICE OF DECOSBIISSIONINGMETHOD AND DESCRIPTION OFACTIVITIES The number and titles of key functions / positions shown on Figure 2-13 may be modified during the course of the decommissioning project. However, the following key functions / positions will not be eliminated while decommissioning activities are in progress, without prior NRC approval:
Waltz Mill Site Manager Radiation Safety Officer Remediation Team Program Manager Radiation Safety Committee 2.4.1 Procedures Decommissioning activities will be performed in accordance with written procedures and guidelines. Procedures will be controlled, prepared, reviewed, revised, approved, and implemented to ensure that operations are performed in a safe manner.
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CHOICE OFDECOMMISSIONINGMETHOD AND DESCRIPTION OFACTIVITIES 2.5 CONTRACTOR ASSISTANCE
. Westinghouse management has selected a team of qualified contractors to perform the WTR Decommissioning project. The team consists of Westinghouse-Nuclear Services Division (NSD), Morrison-Knudsen, and GTS-Duratek (formerly SEG). Westinghouse-NSD will be in
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charge of the overall project management and engineering; Morrison-Knudsen will manage the craft laborers who will do the physical work; and GTS-Duratek is responsible for Health Physics support, radiation surveys, and waste packaging, processing, and shipping. Other contractors may be added to the team as-needed throughout the project.
Contractors and subcontractors performing work under this Decommissioning Plan will be required to comply with the applicable Waltz Mill site policies and procedures.
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l 2-19 REVISION 0
CH01CE OF DECOMMISSIONINGMETHOD AND DESCRIPTION OFA CTIVITIES l
2.6 TRAINING PROGRAM Individuals (employees, cowactors, and visitors) who require access to the work areas or a radiologically restricted area will receive training commensurate with the potential hazards to which they may be exposed.
Radiation protection training will be provided to personnel who will be performing remediation work in radiological areas or handling radioactive materials. The training will l ensure that decommissioning project personnel have sufficient knowledge to perform work
- activities in accordance with the requirements of the radiation protection program and l accomplish ALARA goals and objectives. The principle objective of the training program is to ensure that personnel understand the responsibilities and the required techniques for safe handling of radioactive materials and for minimizing exposure to radiation.
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' Records of training will be maintained which include trainees name, date of training, type of training, test results, authorization for protective equipment use, and instmetor's name.
Radiation protection training will provide the necessary information for workers to implement sound radiation protection practices. The following are examples of the training programs applicable to remediation activities.
2.6.1 General Site Training A general training program designed to provide orientation to project personnel and meet the i requirements of 10 CFR 19 will be implemented. General Site Training (GST) will be '
required for all personnel assigned on a regular basis to the decommissioning project. This training will include: ,
Project orientation / access control Introduction :o radiation protection l Quality assurance i Industrial safety Emergency procedures 2.6.2 Radiation Worker Trainine Radiation Worker Training (RWT) will be required fer decommissioning project personnel working in restricted areas and will be commensurate with the duties and responsibilities being performed. Personnel completing RWT will be required to pass a written examination on the material presented. Completion of this training will qualify an individual for unescorted access to radiologically controlled areas. RWT will include the following topics:
Fundamentals of radiation 2-20 REVISION 0
CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES Biological effects of radiation External radiation exposure limits and controls Internal radiation exposure limits and controls Contamination limits and controls Management and control of radioactive waste, including waste minimization practices Response to emergencies Worker rights and responsibilities In addition to a presentation of the topics identified above, participants in RWT will be required to participate in the following demonstratior.s:
The proper procedures for donning and removing a complete set of protective clothing (excluding respiratory protection equipment)
The ability to read and interpret self-reading and/or electronic dosimeters !
The proper procedures for entering and exiting a contaminated area, including use of l
proper frisking techniques An understanding of the use of a Radiation Work Permit (RWP) by working within the requirements of a given RWP Personnel who have documented equivalent RWT from another site may be waived from taking training except for training on Waltz Mill administrative limits and emergency response, and will be required to pass the written examination and demonstration exercises.
2.6.3 Respiratory Protection Trainine
, Individuals whose work assignments require the use of respiratory protection devices will receive respiratory protection training in the devices and techniques that they will be required to use. The training program will comply with the requirements of 10 CFR 20 Subpan H, 4
Regulatory Guide 8.15 (Ref. 2), NUREG-0041 (Ref. 3) and 29 CFR 1910.134. Training will consist of a lecture session and a simulated work session. Personnel who have documented equivalent respiratory protection training may be waived from this training.
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2-21 REVISION 0 l
. Cil01CE OFDECOMMISSIONINGMETIIOD AND DESCRIPTION OFACTIVITIES 1
l 2.7 OPTIONAL DECONTAMINATION AND DISMANTLEMENT ACTIVITIES WITHIN THE WTR CONTAINMENT BUILDING
, In addition to removal of the reactor vessel internal contents, the reactor vessel, and the
) biological shield, decontamination and dismantlement activities may be performed in other
, areas within the WTR containment building. These activities are not required for TR-2 i
] decommissioning; however, they may be performed prior to transfer of remaining residual a radioactivity to the SNM-770 License.
4 j The decontamination techniques and methods described in Sections 2.2.3.2 through 2.2.3.4, !
- and the dismantlement techniques described in Section 2.2.3.1 may be used to decontaminate and dismantle equipment and structures in these areas.
- These optional activities are discussed as follows
i i, 2.7.1 Sub-pile Room i General l The sub-pile room is a 15' x 15' room located below the reactor vessel. This room has a %-
) inch steel liner on all four walls covering the concrete biological shield. The floor is uncoated
- concrete. The WTR canal mns through the sub-pile room (north-south), separating the room i into two areas (east and west). The two doors to the sub-pile room consist of a steel liner j filled with 12 inches of poured lead. One permits accesses to the east side of the canal and the j other to the west side. The WTR fuel chute is accessible in the northeast corner of the room l
! through a shielded opening in the fuel chute pipe chase. The sub-pile room contains primary l system piping, rabbit tubes, test loop piping and instrumentation piping.
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i Internal Eauipment l The sub-pile room will be cleaned and all remaining piping will be dismantled and/or cut-out
- in disposable sized sections and removed.
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! Floor and Walls l Following removal of remaining loose debris, the area will be re-surveyed for loose and fixed j radioactive contamination to determine the appropriate floor, wall and ceiling surface i decontamination method. Destructive methods such as scabbling, full or partial demolition i may be performed. If scabbling equipment incorporates a self-contained ventilation and
- filtration system (HEPA), additional containment of the work area may not be required.
j Demolition of the intact stmeture may be performed with the demolition of the biological 3
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2-22 REVISION 0 4
CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OF A CTIVITIES l
1 2.7.2 Rabbit Pump Room 1
General The Rabbit Pump Room measures approximately 6'6" by 10'0" by 7'6" high and is located on l
the operating floor along the north wall of the containment building. The Rabbit Pump Room !
contains pumps and valves that delivered the rabbits (test material samples) in a container, to the reactor core via the rabbit tubes. '
Internal Eauipment I Decommissioning activities within the Rabbit Pump Room consist of the dismantling and I removal of the pumps, valves, piping and control assemblies. To control airborne radioactive ,
contamination, the Rabbit Pump Room may be contained and fitted with a HEPA filtration I system capable of creating a negative pressure within the room during equipment removal ;
operations.
Floor. Walls and Ceiline I Following removal of remaining loose debris, the area will be re-surveyed for loose and fixed radioactive contamination to determine the appropriate floor, wall and ceiling surface decontamination method. Destructive methods such as scabbling, full or partial demolition may be performed. If scabbling equipment incorporates a self contained ventilation and filtration system (HEPA), additional containment of the work area may not be required. ,
Demolition of the intact structure may be performed with the demolition of the biological l shield. l 2.7.3 Test Loon Cubicles General Three test loop cubicles are located along the west side of the reactor vessel adjacent to the ,
reactor biological shield. Each cubicle is constructed of concrete of varying dimensions and all cubicles are currently vacant. )
i Floors. Walls and Ceihnes l Fixed and transferable contamination is found on the cubicle floors and fixed contamination on j the walls and ceilings of the cubicle. Following removal of loose debris, destructive methods such as scabbling, full or partial demolition may be performed. Additional containment may l not be necessary if equipment utilizes self contained filtration and ventilation.
Demolition of the intact cubicle stmetures may be performed with the demolition of the l biological shield I l
2-23 REVISION 0 1
CIl0 ICE OF DECOMMISSIONING METIlOD AND DESCRIPTION OF A CTIVITIES 2.7.4 Test Iron Dump Tank Pits General Two 8'0" by 9'0" by 13'0" high Test Loop Dump Tank Pits are located below the operating floor on the east and west side of the transfer canal below the reactor vessel. Ti e west tank pit contains three steel tanks approximately 12' tall and 4' in diameter. The east pit is flooded with water and the pit interior currently inaccessible.
Tank (and Pit) Water Removal The water in the flooded east pit will be pumped out and routed to either the site radioactive water processing facility or treated with a portable system. Once the water in the pit has been removed, the floors, walls, tanks, and tank internals will be surveyed for fixed and transferable contamination.
Any remaining water in the tanks located in the east and west pits will be pumped out to either 4
the site radioactive water processing facility or treated using a ponable system.
All liquids removed and treated from both tanks and the flooded pits will be sampled and analyzed for determination of the proper disposal mechanism.
Tank Demolition Following removal of the water within the tanks, the tanks will be removed from the pits intact, decontaminated and cut into appropriately sized dimensions for packaging and disposal.
The tanks may also be shipped off-site intact following decontamination should a satisfactory salvage opponunity be identified, or shipped off-site intact to a licensed waste processor.
Pit Areas Fixed and transferable radioactive contamination has been found on the floor and walls (and exposed area of the concrete shield plugs) and is also anticipated to be found on the surfaces of 4
the flooded pit once the water has been removed. Destructive methods such as scabbling, full or partial demolition may be perfonned. The pits may be left in place.
Duct Decontamination and/or Removal Supply and exhaust air ducts may be decontaminated in place or removed, decontaminated and released. If warranted by radiological conditions, a temporary HEPA filtration system may be attached to the ventilation ducts to ensure that any loose contamination is drawn away from workers during these operations. Contaminated ducts which can not be decontaminated efficiently and economically may be removed, cut and sized for packaging and disposal.
Resulting penetrations through the containment building will be sealed.
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4 2-24 REVISION 0
. . - - - ~ . - - - - - - . - - - - - - - - . - . . - . - . . - - . - -
F l CHolCE OFDECOMMISSIONINGMETHOD AND DESCRIPTION OFACTIVITIES i 2.7.5 Utilities i
- General Prior to removing electrical, service water, service air, fire or HVAC systems, each. system will be inspected by a qualified individual. All efforts will be made to review the existing status of the respective utilities with Westinghouse service personnel who have a working knowledge of the utilities to prevent service disruption to other site facilities. Emergency utilities, such as fire alarm systems, will be maintained, as required.
Utility Removal Initially, electrical systems will be disconnected. Piping systems will then be removed in areas where electrical systems have been disconnected / removed. Reactor and containment building utilities can be removed simultaneously. As lines are disconnected, provisions will allow the
! collection of any remaining fluid.
Characterization results indicate the presence of fixed and transfera$le contamination on some portions of electrical wiring, conduit, cable trays, electrical boxes and piping. As the utility i systems are removed, contaminated piping, conduit, cables, etc., will be separated from non-
.. contaminated systems.' Fluids collected from piping systems will be sampled and analyzed for i determination of the proper disposal method.
It is expected that dry and/or wet wiping techniques will be sufficient to decontaminate those portions of the materials initially found contaminated. Contaminated materials which can not be successfully decontaminated for unrestricted release or if decontamination is not feasible or cost effective can be volume reduced to the maximum extent practicable and packaged for disposal. Clean material will be disposed of at a local landfill or recycled, if appropriate.
2.7.6 Primary Coolant Pine Tunnels !
General The primary coolant pipe tunnels surround the north end of the transfer canal along the east l and west sides of the reactor vessel below the operating floor. Each tunnel measures approximately 5'0" wide by 10'0" high by 39'0" long and merge into a common tunnel at the ;
north side of the containment building. The tunnel continues below grade to the northeast to i the Facilities Operations Building.' The pipe tunnels contain the primary coolant circulation supply and return lines, demineralizer, emergency coolant and various other piping systems.
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~ Pine Removal Initial work in the pipe tunnels consists of installing temporary lighting and a HEPA filtration system. Identification and tagging of all piping systems will be performed prior to any pipe removal. Any water contained within the tunnels will be pumped out to either the Waltz Mill radioactive . water processing facility or treated using a portable system. The piping components identified as contaminated will be dismantled and/or cut into disposal dimensions 2-25 REVISION 0 i
s CHOICE OFDECOMMISSIONINGMETHOD AND DESCRIPTION OFACTIVITIES or separated for decontamination. Pipe ends should be wrapped as they are dismantled / cut.
The processed or treated liquids will be sampled and analyzed to determine the proper disposal 4
mechanism.
j Tunnel Floors. Walls and Ceilines j' Following removal of all contaminated piping systems, the pipe tunnels floor, wall and ceiling concrete surfaces will be surveyed to determine the extent of contamination. Following
' t removal of loose debris, radiological conditions will determine the appropriate floor, wall and *
- ceiling surface decontamination method for the tunnels. Destructive methods such as scabbling, full or partial demolition may be performed. The tunnels may be left in place.
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2.7.7 Transfer Canal '
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- i. General 1
The transfer canal is approximately 19 feet deep, varies in width from 7 feet to 10 feet and is approximately 160 feet long nonh-south down the axis of the reactor. The canal begins north of the biological shield and continues beneath the reactor vessel to the south, through the G
{. building Annex, ending beneath the Hot Cell area. The transfer canal was the means of
! transporting spent fuel rods from the reactor vessel to the Annex Building and irradiated test
( specimens to the Hot Cell area (see Figure 2-14). The fuel rod conveyor, storage racks, !
l thimble loading machine, transfer chute, rabbit tubes, piping and pipe supports were left in the canal following the 1962 shut down. All irradiated material was removed and properly i dispositioned.
l Exterior Eauipment and Materials l The initial phase of canal remediation will require removal of exterior appurtenances above the l canal (between the 15' and 19' elevations). This may include removal and decontamination of l the drive mechanisms, platforms and existing wire mesh and foam covers.
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, Sediment Removal
! The transfer canal has sediment attached to the walls, floor and structural debris system, in j addition, the concrete sealant is peeling off. This sediment is generally contaminated and in some locations highly contaminated. A filter system will be designed to remove, safely j contain the sediment, and shield workers prior to or during lowering the. water. Figure 2-15 l illustrates one of many ways to accomplish the removal of sediment.
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Canal Water Removal and Interior Wall Surfaces
- - The existing water within the canal will be pumped through a water filtration system to remove
[ fine particles suspended in water. After the water is cleared of solids, the existing or a
- supplemental liquid radioactive waste treatment system will be utilized to treat the canal water.
[ The water level can be lowered in stages and the walls cleaned, as required, from a platform suspended from the canal walls at elevation 19', Figure 2-16. This method will not allow l large ponions of contaminated surfaces to be exposed above the water level. As water is
- removed from the canal, the radiological conditions within the canal will be monitored.
i 2-26 REVISION 0 a
, - - . , _ . , . p -m e , , - - - - - --m- -
CHOICE OFDECOMMISSIONINGMETIIOD AND DESCRIPTION OFA CTIVITIES Appropriate precautions will be taken to prevent or minimize the potential for airborne radioactive contamination. This may include containment and HEPA filtration, maintaining contaminated surfaces wet, or both. Destructive methods, such as scabbling, may be required to remove the fixed contamination on the walls.
2.7.8 Containment Buildine The WTR was a low pressure, low temperature, water cooled 60 MWt reactor housed in a cylindrical vapor containment building. There are two airlocks, and a large overhead door that provides access from the truck lock to the WTR. A schematic of the WTR is shown on Figure 2-1.
The reactor core support stmeture is 29 feet in diameter and 36 feet tall, which houses the reactor pressure vessel. The biological shield surrounding the reactor vessel is made of )
l magnetite bearing concrete, is a total of 44 feet in height and is up to eight feet thick from the j 32 to 51 foot elevations. The operating floor is on the 16 foot elevation and is constructed of l
concrete. The containment is 90 feet in diameter, with a total floor area of 5000 square feet. l There are four support platforms: the truck lock, the reactor head stand, reactor head, and the beam port platforms. As part of the materials testing that was included in the WTR's operational charter, there were several controlled environment test loops installed in concrete cubicles and in an underground test loop vault. Since the shut down most of these loops have been removed.
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The containment building also houses the rabbit pump room, polar crane, and other support systems such as: piping, electrical conduit and boxes, plant and instrument air lines, hydraulic lines, steam and condensate lines, and ventilation ductwork.
Decontamination of the interior of the structure will be conducted only after all other major components have been removed or addressed. Decontamination of the structure will use non-destructive methods if it is to be left in place. If the structure will be removed completely, then it will be shipped to a licensed scrap metal processing facility according to their license requirements. All remaining piping, platforms, and ductwork will be dismantled and either l cleaned, and free released, or sent off site to a licensed waste processor. The cleanup or dismantling should proceed from the top of the dome downward using erected staging or power manlifts for access. Finally, the surface of the operating floor will be decontaminated.
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2-27 REVISION 0
- CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES REFERENCES FOR SECTION 2 l 1.
" Westinghouse Electric Corporation, Waltz Mill Facility, Characterization Report; ;
Nuclear Material License TR-2, Test Reactor License," Scientific Ecology Group, {
- dated February 1994.
- 2. NRC Regulatory Guide 8.15, " Acceptable Programs for Respiratory Protection,"
.l October 1976.
- 3. NUREG-0041, " Manual of Respiratory Protection Against Airborne Radioactive
- Materials," October 1976. ;
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2-28 REVISION 0 l l
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3 j CH01CE OFDECOMMISSIONINGMETHOD AND DESCRIPTION OFACTIVITIES f Table 2-1 l WTR FACILITIES, DECOMMISSIONING ACTIVITIES AND ESTIMATED i
WORKER EXPOSURE i i
1 l WTR FACILITY AREA PROPOSED DECOMMISSIONING ESTIMATED !
j ACTIVITIES l EXPOSU'RE
! (Person-rem)
Pre-decommissioning Activities l Establish radiological controls. 0.05 i Reactor Vessel, Internal Remove internal contents. Use a diamond wire 26.14 Contents, and Biological Shield saw to section the biological shield into slabs and I
- section reactor vessel. (Option 2) l 3 -
Ih Sub-pile Room * ' l Components removed, concrete decontamination, 0.85 1 l and partial or full demolition.
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- Rabbit Pump Room
- l Components removed, concrete decontamination, 0.08 I j l l and partial or full demolition. ;
Test Loop Cubicles m l Components removed, concrete decontamination, 0.13 l and partial or full demolition.
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{ Test Loop Dump Tank Pits
- Compon:nts removed, concrete decontamination, 0.28
- and partial or full demolition.
l l Primary Coolant Pipe Tunnel l Piping removed, concrete decontamination, and , 1.88 l l l partial or full demolition.
! Transfer Canal
- l Water drained, sediment removed, concrete 7.93 l decontaminated, and partial or full demolition.
j Vapor Containment Building Miscellaneous systems and components 0.89
! and Misc. Systems and decontaminated and/or removed, concrete and i Components ** structure surfaces decontaminated, and polar l crane decontaminated.
- ' TOTAL 38.23 l i 1 I
(" The total exposure estimate for the one piece reactor vessel removal, internal component removal, and
- biological shield sectioning and removal (Option 1) is 18.25 person-rem.
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- Decommissioning of these and other structures may be undertaken as part of the WTR Decommissiomng i j project, and will be completed in conjunction with remediation of SNM-770 facilities. !
- See Table 2-1(a) for complete list of miscellaneous systems and components.
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L i 2-29 REVISION 0
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M CHOICE OFDECOMMISSIONINGMETHOD AND DESCRIPTION OFACTIVITIES 4
1 TABLE 2-1(A) 4 LIST OF TR-2 MISCELLANEOUS SYSTEM AND COMPONFNTS CONSIDERED
=
a Transfer Building Pool HVAC Ducts (2)
Experimental Cooling Water
- LLRW Liquid Drain Process Vent i Electrical Cond & Boxes Plant & Instrument Air
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Dionized Water
.i Steam & Condensate Lines Polar Crane Contaimnent Building !
Final Surveys Operating Floor 16' Elev Truck Lock Platform i 1
Beam Port Platform WTR Head Stand Platform WTR Head Platform l
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CIIOICE OFDECOMMISSIONING METIIOD AND DESCRIPTION OFA CTIVITIES FIGURE 2-1 ACCESS CONTROL POINTS TO WTR BIOLOGICAL $H. ELD RGROUND ENTLATION DUCTS TP FACRJTES OPERATK)N BLDG.
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. ..... .. .. ................j_l. .. .. .....
Bot 1/ CANAL EL. ."=V
,,,_.; m._ m,ye x,_,_
i 2-32 REVISION 0 l
CHOICE OFDECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES FIGURE 2-2 CONCEPTUAL DRAWING OF RAIL TRANSPORT SYSTEM TO REMOVE MATERIAL FROM WTR i
l CRANE TRUCK i
REACTOR i
RAIL FOR CART '
4 EXTEND RAIL 10'-0" BEYOND 1 TENTED AREA g
... z.......u ...... 1 .
} -
RAIL C RT
'20' LC. X 15' WIDE O LOAD TRUCK LOCK BAY
! TENTED AREA AREA l!
- TRUCK LOCK PLATFORM STAG]NG AREA ,,
FOR OFF LOAD ! l-OF CONTAMINATED !i 4 AND l NON-CONTAMINATED '
l WASTE F HEPA ,
FILTER ;
i 12'- 0" l
l l
l t
2-33 REVISION 0
CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES FIGURE 2-3 CONCEPTUAL PLAN HAUL CHART I
--= '
F:-, -2. I
- 6"_~._._ i
% i==-3 e-:::
m .. .
ur
, -t" "i -'
I I l 1
, {. . f.- . _.
i
/
I x
_- .., s '
/> i--w , 4----i, t
\\ i 1--H 6 =1 m ,-
I CONCEPTUAL PLAN HAUL CART
\ ,
~g .,
4 \
_ L _ 4 \
2-34 REVISION 0
CHOICE OF DECOMMISSIONING METIIOD AND DESCRIPTION OFA CTIVITIES FIGURE 2-4 (sht.1 of 3)
CONCEPTUAL DRAWING
., ONE PIECE REACTOR VESSEL REMOVAL 4
N
)
4 i
l.
. REACTOR VESSEL j W/ SHIELDING
. EST. WT.
I 286,000 POUNDS J
4 i.
4'-2" l
l.
i l 74' 1 i.
. - - - I hee CC' a 70' i o !
~30' 35' 1
65*
2 REVISION 0 l
CHOICE OFDECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES FIGURE 2-4 (Sht. 2 of 3)
AREA PLAN FOR ONE PIECE REMOVAL FROM CONTAINMENT l
I l
CONTAINMENT i
TR-2 REACTOR WITH l CONCRETE SHIELDING
[___
e65' LIFT RAD.
TRUCK 70' '
LOCK BAY
@@ i'
\ STEP NO.1
40,
/ REMOVE REACTOR AND
)
s SET OUTSIDE AT 32', R j A ~
g7 f Wl,
/ 32'R g
1000 TON 43' YMM LRANE POSITION #1 u
1 2-36 REVISION 0
CHOICE OFDECOMAtlSSIONING METIIOD AND DESCRIPTION OFACTIVITIES FIGURE 2-4 (Sht. 3 of 3)
AREA PLAN FOR ONE PIECE REMOVAL / LOADING OUT TO TRUCK CONTAINMENT TR-2 REACTOR EACTOR ASSY STEP #2
'OR LOADING -
F ONTO TRUCK TRUCK LOCK
)
/
\g a -
g' gREACTOR ASSY i
b/ =
40' -
/ D&
l DE y
/
/j 1 [
62' CRANE's
\
Q '
A-LIFT '
65' RAD.
-CRANEf ESN ' t RAD. W '
)O TON t' 4' "G" BUILDING LNE POSITION #2 2-37 REVISION 0
CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES FIGURE 2-5 i l
ELEVATIONS I
REMOVE ALL INTERiiAL COMPONENTS POSSIBLE
' l ABOVE THE CORE WITH THE OVERHEAD CRANE.
4 E'.EV. 61'-8tn
' l i
l / I l i 'l I/ r l'
i ELEV 51.0*
i
. l
_ ELEV 36.6' l
ELEV 32.25' E
\
4 -
l ELEV 19.0" ELEV 16.0*
umumme ELEV 0.0' REACTOR SECTION 2-38 REVISION 0
. CIJOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES \
FIGURE 2 6 UPPER PORTION REMOVAL OF BIOSHIELD BY SECTIONING fv l f [$ _ REMOVE THE UPPER PLATFORM AND UPPER 19' DI A. SECTION OF THE B10 SHIELD BY t %pg ~.q ;+*-p, e 2, N1 CROSS CORING AT 3* INTERVALS.THIS WIll ALLOW THE SECTIONS OF BIOSHIElD TO
+ <w
- d It?ljY BE DI AMOND WIRE SAW Cut INTO 20 TON Ig'f F ' BLOCKS FOR REMOVAL WITH Tete OVERHEAD M
O. %fL p:'f@F -(ig : ;
- REWOVAL OF THE TOP LAYER OF BIOSHE!LD isr y'
%bp. g/ WILL REQUIRE 66 LINEAR FEET OF CORE DRILLING AND 680 50. FEET OF DIAMOND WIRE SAWING.
l x ._. .
/ c55 'oc^n o"5 14 _., ELEY51.0*
_ _- ._!_J _ , . -
/ ,
1 t
_ ea, u EtEv 3m j s
y ,
s
+
111l ll i
\ , , /
l i ELEv to 0-l ELEV 16.0'
! l l
, I r
< )
g i
I Euv 0.0' UPPER POR ON REMOVAL OF :
RrAcTOR SFCTION BIOSHIELD BY SECTIONING '
(NOT TO SCALE) 2-39 REVISION 0
CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES FIGURE 2-7 MID PORTION REMOVAL OF BIOSHIELD - SECTION M PLAN AT RIGHT ILLilSTRATES M CORE ORILL PATTDIN UTIL.12ED TO RDdOW THE NOTE: AN ADDITIONAL 1* SECTION WILL BIOSHI AWAY FROM THE CDCER CORE. BE RDdOG FROM THE INNER PORTION M AND 8 WOLED BE RDIOC BY INDIVIDUALLY AS MY
.(A) _ y Or THE RDeos BLOOG DUE TO LOW LENEL ACTIVATION IN THE B105HIELD (g)l NEAR THE CX)RE.
g g a gi rREE THg i n N g LEA E' APPRQK1WATELT 1 CONCRETE STADOINC , (s)\. l AFB RDM FROM M CmR MdN)
,_ . _ - _ ..u. .a
, CROSS CORE LOCATION
_- . - . - l i
(A) _. ..
_.t . . . . . . . _ .
, (B) i I
(A)
DUE TO WEIGHT RESTRICTION THE '~#^' nrV 51.0' CORE DRILL PATTERN AND 5AW CUTTING PATTERN WILL MAVE TO l
1 BE REPEATED AT 2M INTERVAL 5. -
l AN APPRCXIMATE TOTAL OF 700' OF CCRING AND 4.416 5Q. FT. OF ~ ' ~
DIAMOND SAW CUTTING TO REDUCE _ l _
THE BIOSHIELD PARAMETERS DOWN i fo nEv. r.0. _
i i _
nrvss.s*
nrv 32.25' l ~= -7 .
l g
, l .__..
II Il ll
\ i., /
- nEvis.0- I aEvis.r 4
___________\___ ___ _______
i l
,l ll l.l ll
'r \ ,q/
TRANSFER CHUTE /g ;
.; .s.
ELEV 19 0'
./,,
! ELEV 16.0' //
/*
ll l'
l
-l-----.------ ------------
L h nEv o.0' l
)
~
REMOVAL OF LOWER B10 SHIELD 4 amm mTTaN SECTION (NOT TD $CALE) i 2-42 REVISION 0 1 2
- . . . . .- . _. - . ..~
l CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES FIGURE 2-10 REMOVAL OF LOWER INTERNALS TRUCK LOCK PLATFORW WILL BE RE8UILT FOR FUTURE USE.
, _ , _ _ , . - ~ . - _ . . _ _ . _ ,
s l l
.' P
.x 1 0;*J g
/ 'l s I
[N,4
- 1 l . l g l _ _
\ - _-' \ l .} .. f}
- l ,
A.
I i -. - - , - . , . _ . , . _
s /
l I
gtgy, 37y REMOVE THE REMAINDER OF INTERNALS REWOVE THE TRUCK LOCK PLATFORh (SEE OPTION)
I: .I /
= = = = ,
\
I, _,.-
l, j n .q i . e i
TRANSFER CHUTE ,#I _; !"
i l ELEV 19.0*
}
ELEV 16.0' l
'l l
1-l
_ ____ _ _ _ _ . ['_ _____
_ _ _ _ _ _\_ _ _ _ _________
l ELIV 0.0*
~
REMOVAL OF RrAcToR steTTow LOWERINTERNALS OPTION: IF THE REW AINDER Or THE B10 SHIELD CAN BE DECONTAWINAlED WITHOUT FURTHER DISASEMB'.Y. LEAVE THE TRUCK LOCK PLATFORM IN (NOT TO SCALE) PLACE AND REPLACE THE CENTER SECTION CAVITY WITH A STRUCTURAL PLUG 2-43 REVISION 0
CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES FIGURE 2-11 LOWER REACTOR BIOSHIELD SECTION REMOVAL TauCx LOCx HATrORM WII.L BE REBUILT FOR FUTURE USE. 1I -
t,I e lit i
/ ll lll ii ll
'i h
i ,d,- - ,- -_-l,ll_ _ ,f! __ __ _ _ . - _
-~ c,*% ,
% lll ll i jt
'D0%Ji 4
i to
-,A i si t i i
, b:::-:o :- : h,si ::
l $-
L.- h--- 1'II
'l i t l'
11 Il
\ i I
i
/
<^r g CORE LOCATIONS '
ELEY.32'3" REWOVE THE SOUARE SECTION OF THE i 810SHIEl.D ABOVE THE SUBPILE ROOM i/f'
~ ~
~ ' ~ j/ AS SHOWN. 24 BLOCKS WILL BE REQb EACH BLOCK WILL WElGH APPROX.19
~~~~
8 e~ TONS. 200' 0F CORE ORILLING ANO
~,'/f 1328' 0F SAW CUTTING WILL BE REQ'D.
u- -
TRANsrER CauTE // I ELIV 19.O'
,#/ s I ELEV 16.0' ,/
l llll/----------\ ------------
I '
ELEv 0.0' I
~
LOWER REACTOR BIOSHIELD REACTOR SECTTON SECTION REMOVAL NOTE: Ir OPT 10N DESCRIBED IN F10URE 8 IS NOT UTILIZED CONTINUE AS $HOWN. i (NOT TO SCALE) I i
l i
2-44 REVISION 0
CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES FIGURE 2-12 WTR DECOMMISSIONING SCHEDULE kn
~t WTR Deconwnmaaonsng l Schedule
{ f 1996 1997 1990 1999 2000 2001 2002 2003 2004 2005 2006 n Task Warne Duret6on Start Finesh 1 Protect Prop...iiv.,($NM-770
- TR-2) 196d Tue 10/15!96 Tuo 7/1S/97 '
M5 E 7q5 !
I ' ' '
{
3 Remon Structures and interferences (SNM-770) 121d
, , 4
- l l Tue 7/15/97 Tue 12/30/97 7/15 E 12/30 4 i i j l i
& Remedete Contanunated Sod 261d Thu 1/1/98 Thu 12/31/98 gg ggg I
j , b i y !
Rernedste Rebred Facdibes $22d Thu 1/1/98 Fn 12/3149 1/1 12/31
~
9 Submd Concehal WTR Decommasoning Plan 1d Mon 4/7S7 Mon 4/7197 g
l l l l l !
11 NRC Conceptual Decommaserung Plan Concunence 1d Mon 6/30/97 Mon 6/30/97
~
,gg ; i f !
12 ! -
i ! !
13 Subm WrRDecommessemngPlan '
1d Wed 7130/97 Wed 7/30/97 g yg i 14 '
, I 4
l ?
l l NRC Decommesserung Plan Approval Id is- Fn 1/3098 Fri1/30/98 ! g gg 16 '
17 Remove Reactor Vessel /Belogical She64 102tki Mon 2/2/98 Fn 12/28/01 2/2 11/28 18 '
i l l e
19 Waste e ckagmg/Shsppmg 1150d Mon 2/2/98 Fn 6/28/02 6/28 a !' 2,1! ! l 21 WTR LJoense Termination / Transfer to SNM-770 1171d Fn 1/1/99 Fn 6/27/03 l l l j in l 612 7 i l Task M Summary Rolled Up Progress N y Progress N Rolled Up Task l M_ + -o,Mmtone o 1 2-45 REVISION 0
1 i
CHOICE OFDECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES l l
FIGURE 2-13 I WTR DECOMMISSIONING PROJECT RESPONSIBILITY MATRIX 4 ;
i ESBU Dmsion Regulatory /
Quality Systems General Environmental Assessment l Manager Affairs !
)
I Radiation Safety l Comminee '
I Radiation
) Waltz Mill Site Manager (NOTE) a 1
Waltz Mdl Site Environmental Remediation Team Quahty Operations Affairs
- Program Assurance- Manager Project Director Manager WM Sne Operauons Expeditor
! I I Security Safety Radiation .. Westinghouse Mornson GTS Duratek Facilines & Safety Project Knudsen Health Access Manager Environmental ofncer Techmcal Operanons Physics Conut! Lead ' " " (NOTE) Lead Mar:ager Manager NOTE: The WM Radiation Safety OfDcer reports to the WM Site Operations Manager and as also the Secretary of the Radiation Safety Commatee.
apn 2 46 REVISION 0
CHolCE OFDECOMMISSIONING METHOD AND DESCRIPTION OFACTIVITIES l FIGURE 214 !
LOCATION OF TRANSFER CANAL 1
4 1
I
)
UNDERGROUND VElfrt.ATION DUCTS <
BIOLOGICAL SHIELD f YnFACUTESENGDEERINGas. DING l m;V eur mau,iwa noeth .
. /~' esove ca.aa 1 70 FACa. files a==% $cu j,,'yQ cananaa m m sta m d OPERATIONS KDG.
~, j --].f 1 __
. .. y J !
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s k,,--- l ai \ aNNEm m i
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^r .a.me.yan AT ELEVATION 20'.0" i V enoncas *
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l cowf amem eUILDENG 4
, , , , , , r L J pu era - vo.or casat wan et. ira e*T mca .,
tom ooon s s
re.=.A 4+--l / waren tavnire e ,a T ' a- -f
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M ?; y v ........ ....... _ ...............a L C - -...-..-
TRANSTER CANAL-* -- -
-TRANSFER CANAL
/ m sECTm x e sEcf m aD ED N \ oTTN D Ma m iiw,,,en ' ./ ma ss l SECTION @
oes noau
.=== L i
i 2 47 REVISION 0 1
CHOICE OFDECOMMISSIONINGMETIIOD AND DESCRIPTION OFACTIVITIES FIGURE 2-15 TRANSFER CANAL DECONTAMINATION SEDIMENT REMOVAL CONCEPT CONTAINENT WmiMPA SYSTEM TO DRAW NEGATIVE AR
< t d'01A.SCREEEN S CLO RETURN LINE l PUMP SUCTION
-,..__.J. _ _ _ ,
I WASTE CONTAINER f '-
C I
c:w :-x , ,-
L 1
S0 CPM PUMP L. oEte
%./
FIRST STAGE OF CANAL DECONTAMINATION REMOVE SEDIENT & DEBRIS FROM WALLS AND R00R OF THE CANALUT1UDNG AWET VACUUM SYSTEM.THE SOLDS & SEDIENT WILL SE SCREENED OFF ASIT PASSES OVER A COVERED SCREEN OR OTHERFLTER MEDIA ANDTHE FLTERED WATER RETURNEDTO TRANSFIR CANAL THE CANAL.
CRAPER BLADE SEDIMENT IS SCRAPED FROM WAllJ AND RS MD EmR$ WE EN bM SUCTION INTAKC NM TO BE PASSEDOVER WE SCREENS.
N0ZZLE DETAI @
v ,
l l
)
i 2-48 REVISION 0 !
i
CHOICE OF DECOMMISSIONING METHOD AND DESCRIPTION OFA CTIVITIES FIGURE 2-16 TRANSFER CANAL DECONTAMINATION
( IELDED)
SusMtRS m E N ALFUMP
{ emmr suctim O.R ?.$
o < ~
TsammIvat AS THE CANAL W ATER !$ LOWERED THE WATER e W111BE FUMPED THROUGH A MICRO FILTER SYSTIM ANDINTO AHOLDINGTANK THE FILTERS WILL REMOVE THE REMAINING SEDIMEh7 FROM THE g
- WATER AND SHOULD ALLOW THE DISPOSALOF THE L WATERINTOTHE EXISTING DRAIN LINES.
4 M40R 70 RELEASE FROMTHE HOLDNGTANK, ,
THE WATER MUST BE ANALYZED TO ENSURE IT
. hE!ETS REQUIRElENTR FOR DSCHARGEL 4
WA10t REMOVAL SOGATIC 4
HEPA FILTER 1
i i ELEY 16*-0*
,g 7g,
() () /
/ i i SECOND ETACE OF CANAL DECONTAMINATION CANTELEVERED SCAFOLDINO TO BE USED DURING THE REMOVAL OF THE SEDIMENT FROM THE WALLS AND FLOORS.
THE SCAFFOLDING WILL BE LOWERED AS THE WATER LEVEL IS REDUCED AND USED AS A WORK PLATFORM JJST ABOVE THE WATER LEVEL.
TRANSFER CANAL 2-49 REVISION 0
J PROTECTION OF OCCUPA TIONAL AflD PUB 12C HEALTH AND SAFETY SECTION 3 PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY 3.1 FACILITY RADIOLOGICAL STATUS 3.1.1 Facility Operatine History The WTR began operations in 1959, initially operating at a power level of 20 MWt and ultimately operating at 60 MWt. The WTR operated on commercial contracts in which various materials 4
were inserted into the core and removed at the end of a 21-day cycle. Additional capabilities, such as the rabbit facilities, allowed specimens to be inserted into the core and withdrawn during reactor operations, independent of the 21-day cycle.
3.1.2 Cunent Radiological Status of Facility Following final shut down of the WTR in March 1%2, all spent fuel was removed from the site and shipped to Idaho Falls, Idaho for reprocessing. Unirradiated fuel (new fuel in storage) was returned to vendors for cold reprocessing, and irradiated specimens were returned to experimenters or disposed as radioactive waste. The reactor facility was partially dismantled, but not completely decontaminated. Some of the equipment and tooling was left in the transfer canal and reactor internals remained in the pressure vessel. The reactor head was mplaced and secured in accordance with standard procedures. The vessel and prunary coolant system were drained and doors were sealed or secured to prevent unauthorized entry.
During 1993, a complete radiological characterization of the remaining WTR stmetures and components was conducted. The pnmary objective of this effon was to provide sufficient ;
radiological information to develop the WTR Decommissioning Plan and facilitate realistic cost i benefit analysis in support of assessing decontamination and decommissioning options. ;
Numerous measurements and samples were obtained and analyzed to characterize the extent of neutron activition, the radioactive contamination present on the internal surface of components, piping, etc., and the extent of fixed and transferable contamination present on internal and external surfaces of the WTR stmetures and systems. This included the radioactive contamination present in the WTR transfer canal water, in the canal sediment, and on the surfaces of the canal walls and ,
components / materials present in the transfer canal. l The following sections stunmanze the radiological characterization of major WTR structures and systems. This summary identifies the average value for the population of gamma exposure rate, surface contamination (fixed and transferable), material activation, and volume contamination meassurements, where appropriate. A more detailed pmsentation of the radiological data, as well l 3-1 REVISION 0 i
l
PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY I
as the sun'ey and sample collection methodology, characterization, Quality Assurance / Quality '
Controls, and the determination and documentation of background radiation measurements are found in Reference 1 of this Section.
In the following summary of radiological data, "N/A" identifies a specific parameter which was not measured; MDA indicates a value which was below the measurement Minimum Detectable l Activity (MDA). MDA values for the specific measurements are provided in the WTR Characterization Report. Additional systems, not identified in this section, are also included in the WTR Characterization Report. These consist of Electrical Conduit and Boxes, Plant and Instmment Air Lines, Deionized Water Lines, Hydraulic Lines, CRDM Gland Seal Piping, and Steam and Condensate Lines.
3.1.2.1 WTR Stmetures Operating Floor (16' Elevation):
Direct Alpha and Beta Radioactivity (dpm/100 cm2 )
Description Alpha Beta ,
Floor MDA 15,000 Walls @ floor level and 6' N/A 620 Wall Hot Spots N/A 23,000 Transferable Alpha and Beta Radioactivity (dpm/100 cm2 )
Description Alpha Beta Floor 24 550 Perimeter Wall MDA MDA Gamma Exposure Rates (gR/hr)
Description Gamma Floor @ l meter 59 3-2 REVISION 0
PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY Concrete Core Samples Total Gamma Activity (gCi/g) j Description 0.5" Depth 1.0" Depth NW Quadrant 1.7E-6 3.6E-7 NE Quadrant 9.03E-5 9.84E-6 SE Quadrant 3.3E-6 7.9E-7 l
SW Quadrant 3.1 E-6 2.7E-7 Personnel Hatch 8.7E-5 N/A 1.
J The primary contaminant identified by gamma isotopic analysis in all core samples was Cs-137 (Co-60 was identified in cores from the Northeast Quadrant and Personnel Hatch, and accounted for 4% aixi 2%, respectively, of the total activity). I Composite of 0.5" Concrete Core Samples Gamma and Beta Emitting Radionuclides (pCi/ gram)
Cs-137 4.7E-5 Ni-63 1.6E-5 Sr-90 7.6E-5 Total Activity 1.39E-4 Composite of Paint Scraping Samples (pCi/ gram)
Cs-137 7.6E-6 Co-60 3.4E-6 Sub-Pile Room:
Transferable Alpha and Beta Radioactivity (dpm/100 cm 2)
Description Alpha Beta Floor MDA 14,000 Floor beneath east valve 28 20,000 bank 3-3 REVISION 0
PROTECTION OF OCCUPA TIONAL AND PUBLIC HEALTH AND SAFETY Gamma Exposure Rates (mrem /hr)
Description Gamma Floor @ contact <1 Composite of 0.5"and 1.5" Concrete Core Samples Gamma Activity (pCi/g)
Nuclide 0.5" Depth 1.5" Depth Cs-137 2.7E-5 1.7E-6 Co-60 1.1E-5 N/A Sr-90 7.6E-5 N/A U-234 2.7E-5 N/A Pu-238 1.9E-5 N/A Pu-239,240 2.6E-6 N/A Total Activity 1.6E-4 1.7E-5 Composite of Paint Scraping Samples Gamma Activity ( Ci/g)
Cs-137 8.9E-6 Co-60 2.8E-6 Composite of Lead Scrapiag Samples Gamma Activity (pCi/g)
Cs-137 7.3E-6 Co-60 7.9E-6 Rabbit Pump Room:
Direct Alpha and Beta Radioactivity (dpm/100 cm 2)
Description Alpha Beta Floor N/A 34,000 Walls N/A 440 Ceiling N/A 530 3-4 REVISION 0
1 PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY Transferable Alpha and Beta Radioactivity (dpm/100 cm')
Description Alpha Beta Floor MDA 570 i Walls MDA MDA 1
Gamma Exposure Rates (mrem /hr) l Description Gamma Floor @ l meter 17 Concrete Core Samples Gamma Activity (gCi/g)
Nuclide 0.5" Depth 1.0" Depth Cs-137 3.1E-5 3.5E-6
, Co-60 3.5E-6 N/A Test Loop Shield Cubicles:
Direct Alpha and Beta Radioactivity (dpm/100 cm2)
Description Alpha B. eta Gas Loop Test Cubicle Floor MDA 65,U00 Walls MDA 227 West Test Cubicle Floor N/A 240 Walls MDA 235 North Test Cubicle Floor N/A N/A Walls 230 590 3-5 REVISION 0
PROTECTION OF OCCUPA TIONAL AND PUBLIC HEALTH AND SAFETY s
Transferable Alpha and Beta Radioactivity (dpm/100 cm2 )
Description Alpha Beta North Test Cubicle Walls MDA MDA West Test Cubicle Walls MDA MDA :
Gas Test Cubicle Walls MDA MDA J
- Gas Test Cubicle Walls MDA 490
-f Gamma Exposure Rates (gR/hr) ,
Description Ganuna Floor @ contact 86
, Floor @ l meter 18 Concrete Core Samples Gamma Activity (gCi/g)
Description Nuclide 0.5" Depth 1.0" Depth ,
t North Cubicle Walls Cs-137 2.lE-6 2.0E-6
\
Co-60 1.5E.6 1.7E-6
?
Eu-154 N/A 1.5E-6 West Cubicle Walls Cs-137 1.7E-6 1. l E-6 Co-60 2.0E-6 6.9E-7 Gas Test Cubicle Floor Cs-137 3.5E-4 1.4E-6 ,
)
Test Loop Dump Tank Pits:
Direct Alpha and Beta Radioactivity (dpm/100 cm 2)
Description Alpha Beta Walls MDA 67,000 1
l 3-6 REVISION 0 l
)
PROTECTION OF OCCUPA TIONAL AND PUBLIC HEALTH AND SAFETY Transferable Alpha and Beta Radioactivity (dpm/100 cm')
- Description ' Alpha Beta Pit Floor MDA 3,700 Pit Walls 24 1,600 Gamma Exposure Rates (gR/hr)
J Description Ganuna Floor @ 1 meter 99 Concrete Core Samples Gamma At.tivity (pCi/g)
Description Nuclide 0.5" Depth 1.0" Depth Pit Walls Cs-137 1.6E-5 2.5E-6
< Co-60 1.lE-5 2.6E-6 l Pit Floor Cs-137 1.6E-5 8.0E-7 g Co-60 1.2E-5 5.5 E-7 i i
. Truck Lock Platform:
1 Direct Alpha and Beta Radioactivity (dpm/100 cm2) l Description Alpha Beta Floor Tile Surface 94 1,600 l t
Transferable Alpha and Beta Radioactivity (dpm/100 cm2 )
Description Alpha Beta Floor MDA 120 i
4 i
3-7 REVISION 0 l N
~. . .. _ ._ ._- _ ._ . _ . . . _ _ . _ _ _
PROTECTION OF OCCUPATIONAL AND PUBIJC HFALTH AND SAFETY Gamma Exposure Rates (gR/hr)
Description Gamma '
Platform @ 1 meter 13 p East of Bioshield @ 1 meter 13 West of Bioshield @ 1 meter 8 1
a Concrete Core Samples Gamma Activity (pCi/g) .
Description !
Nuclide 0.5" Depth 1.0" Depth ;
Platform Floor Cs-137 3.06E-7 N/A a Beam Port Platform: .
)
{ Direct Alpha and Beta Radioactivity (dpm/100 cm2)
Description Alpha Beta d
Floor Tile Surface N/A 1,200
Transferable Alpha and Beta Radioactivity (dpm/100 cm')
Description Alpha Beta Floor MDA 110 Beam port shield and shield MDA MDA blocks Pump Skid MDA 170 Gamma Exposure Rates (gR/hr)
Description Gamma Platform @ 1 meter 9 Pump Skid @ contact 8 3-8 REVISION 0
l PROTECTION OF OCCUPA TIONAL AND PUBLIC HEALTH AND SAFETY l
Concrete Core Samples Gamma Activity (pCi/g)
Description Nuclide 0.5" Depth 1.0" Depth l Platform Floor Cs-137 2.0E-7 1.5E-6 1
i Co-60 N/A 1.0E-6 !
l i
i l Reactor Head Stand Platform:
- Direct Alpha and Beta Radioactivity (dpm/100 cm2) l Description Alpha Beta Platform & Catwalk N/A 740 l Reactor Head Stand 63 1,200
Transferable Alpha and Beta Radioactivity (dpm/100 cm ) 2 i
Description Alpha Beta i
[ Chase '
Ganuna Exposure Rates (pR/hr)
Description Gamma
}
l Platform @ l meter 7 Head Stand @ contact 12 Bioshield Catwalk @ contact 10 d!
Cable Chase @ contact 8 i
l I
3-9 REVISION 0
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PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY I
Reactor Head Platform: 1 Direct Alpha and Beta Radioactivity (dpm/100 cm2)
Description Alpha Beta Deck Plates 96 860 i Platform Storage Area 98 710 CRDMs 59 5,300 Reactor Head 84 MDA 4
Transferable Alpha and Beta Rcdioactivity (dpm/100 cm 2) :
Description Alpha Beta Deck Plates MDA MDA Platform Storage Area MDA 34 Reactor Head MDA 300 Top of Bioshield MDA 120 -
Composite of Smear Samples Fraction of Total Gamma Activity (pCi/g)
l 3-10 REVISION 0
PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY 3.1.2.2 WTR Systems Rcsactor Vessel and Internals:
Direct Alpha and Beta Radioactivity (dpm/100 cm2 ) l Description Alpha Beta i Reactor Head Access Internal surface 220 86,000 Plugs and Flange External surface MDA 420
. Iower Reactor Vessel Sub-pile platform and canal sides MDA 19,000 Flange Bottom of vessel MDA 7,222 Flanges in sub-pile room N/A MDA Internal Thermal Shield 3" thermal shield 620 N/A 2
Transferable Alpha and Beta Radioactivity (dpm/100 cm2 ) i Description Alpha' Beta Reactor Head Access Plugs - Internal 38 88,000 Plugs and Flange Plugs - External MDA 250 Rx head and CRD Externals MDA 430 l Rx head and CRD Internals MDA 890 Upper Reactor Vessel Vessel walls, 54.5' to 58' MDA 1,700 -
Internals 2,100 l Sides and bottom of CRD 46 3,600 Core plate top MDA 990 Fuel chute MDA 550 IAwer Reactor Vessel Sub-pile platform and canal sides MDA 920 Flange lower vessel internals 19 5,400 Lower Vessel Drain 1" Thermal shield- MDA 240 Internals 2" Thermal shield MDA 430 3" Thermal shield MDA 570 l
3-11 REVISION 0
1 PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY I
Gamma Exposure Rates
- 1. Upper Reactor Vessel Internals A gamma radiation profile was obtained at one foot intervals, starting at the head access pons and l terminating approximately 12 feet into the reactor vessel. Attempts were made to maintain the detector vessel geometry constant for all measurements. ,
Gamma exposure mtes, starting at the reactor head, range from 113 mR/hr to 1800 mR/hr.
Gamma exposure stes were also obtained for the CRD and Fuel chutes. In both cases, the gamma exposure rates dropped to below 100 mR/hr approximately 5 to 6 feet into the chute (CRD chute )
rate dropped to 1 mR/lu-). '
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- 2. Iwwer Reactor Vessel Gamma exposure rates as high as 400 mR/hr were found on some components, with a general area gamma exposure rate of 1-3 mR/hr in the Sub-File Room.
- 3. Reactor Internal Thennal Shields and Core lattice Thermal shield contact gamma exposure rates range from < 0.2 mr/hr to 8,250 mr/hr. Core lattice gamma exposure rates range from 27 R/hr to 62 R/hr. ,
Isotopic Analysis of Composite Smear Samples and Material Scrapings Smear composites, metal scrapings / cores, and samples of miscellaneous materials removed from I the reactor vessel were analyzed to identify the radionuclides present and the rela:ive abundance of each. In all cases, the prunary gamma emitting radionuclides included Cs-137 and Co-60. As 1 expected, analysis results for samples obtained from areas not directly influenced by the core neutron flux indicated that Cs-137 was the predominant gamma emitting radionuclide and for samples from areas which were in clo,e proximity to the core, Co-60 predominated. A trace amount of Sr-90, generally much less than 10% of the total activity, was also identified in a number of the samples.
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3-12 REVISION 0
PROTECTION OF OCCUPA TIONAL AND PUBLIC HEALTH AND SAFETY Biological Shield:
Direct Alpha and Beta Radioactivity (dpm/100 cm 2)
Description Alpha Beta Magnetic Plugs MDA 380 !
Instrument Penetration 77 2,300 Covers (External)
Pipe Sleeve Plugs MDA 2,300 Bioshield @ Top of Reactor MDA 21,000 Transferable Alpha and Beta Radioactivity All average transferable alpha and beta contamination was at or below the detection MDA with the exception of the exterior surface of the instmment penetration covers which had an average beta activity of 76 dpm/100 cm2, Gamma Exposure Rates Gamma radiation profiles were obtained over the penetration length of several instrument penetration tubes at intervals of one foot from the top of the bioshield down to the sub-pile room.
Gamma exposure rates range from < 1 mR/hr to 65 mR/hr Isotopic Analysis of Concrete Cores, Miscellaneous Samples and Material Scrapings Metal scrapings / shavings, concrete cores, and samples of miscellaneous materials removed from the bioshield at various locations were analyzed to identify the radionuclides present and the relative abundance of each. In all cases, the primary gamma emitting radionuclides included Cs-137 and Co-60.
Tratufer Canal:
'The transfer canal remains filled with water, therefore, direct alpha and beta radioactivity measurements were not possible.
Transferable Alpha and Beta Radioactivity (dpm/100 cm2)
Description Alpha Beta Canal Wall, North 5,900 270,000 Canal Wall, Sub-Pile 6,100 330,000 Canal Wall, South 7,000 320,000 Gamma exposure rate profiles were taken along the canal walls at several locations, including the north, south, and sub-pile canal sections. Gamma exposure rates were also taken at specified l
J 3-13 REVISION 0
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PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETl' locations along the canal walls, bottom and on canal components, the rabbit tube and indexing station. The following are typical gamma exposure rates:
Gamma Exposure Rates (mrem /hr)
Description Gamma Canal Wall Top 0-2 Canal Wall Bottom 50 Rabbit Tube Indexing System 30 Canal Bottom 50 Canal Components 50 l Isotopic Analysis of Composite Smear Samples and Material / Sediment Smear composites and samples of various materials from within the canal were analyzed to identify the radionuclides present and the relative abundance of each. In all cases, the primary gamma i
emitting radionuclides were Cs-137 and Co-60, with Co-60 being the predominant contaminant.
" Concrete Core Samples Gamma Activity (gCi/g)
Description Nuclide 0.5" Depth 1.0* Depth Canal Floor Cs-137 1.5E-3 5.0E-3 4
, Co-60 4.8E-5 1.2E-5 Am-241 N/A 3.2E-4 i
} Primary Coolant Piping Systems:
Transferable Alpha and Beta Radioactivity (dpm/100 cm2)
Description Alpha Beta l Piping Exterior MDA 510
- 1. cad Bricks MDA 1,600 Gamma Exposure Rates The average gamma exposure rate for the primary coolant piping B 4,300 R/hr.
i 3-14 REVISION 0
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PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY Isotopic Analysis of Various Material, Sediment Samples Samples of various materials from within the primary coolant pipe tunnel were analyzed to identify the radionuclides present and the relative abundance of each. Samples included water, sediment, scale, lead scrapings, and pipe insulation. In all cases, the primary gamma emitting radionuclides were Cs-137 and Co-60, with Cs-137 being the predominant contaminant.
Sub-Pile Room Thimble Valve Banks:
Transferable Alpha and Beta Radioactivity (dpm/100 cm2 )
Description Alpha Beta Exterior East & West Valve MDA 280 Banks Interior East & West Valve MDA 14,000 fBanks Gamma Exposure Rates The average gamma exposure rate for the west valve bank is 1-3 mR/hr (indistinguishable from the sub-pile room general area gamma exposure rate).
Isotopic Analysis of Composite Smear Samples and Material / Sediment Samples of sediment and metal from the test thimble bank were analyzed to identify the radionuclides present and the relative abundance of each. In all cases, the primary gamma emitting i radionuclides were Cs-137 and Co-60, with Cs-137 being the predominant contaminant for a majority of samples. I l
Rabbit Pump and Piping: !
Transferable Alpha and Beta Radioactivity (dpm/100 cm')
Description Alpha Beta Rabbit Tube Internals 310 210,000 ,
1 Rabbit Pump Internals 19 670 Rabbit Pump Externals MDA MDA l
l Gamma Exposure Rates The average gamma exposure rate for the rabbit pump is 24 R/hr.
Isotopic Analysis of Composite Smear Samples and Material / Sediment Samples of the rabbit tube and from the rabbit pump drain plug were analyzed to identify the radionuclides present and the relative abundance of each. In both cases, the primary gamma mi: ting radionuclides were Cs-137 and Co-60.
3-15 REVISION 0
PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY
- Gas Test Loop Tanks and Piping: '
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Direct Alpha and Beta Radioactivity (dpm/100 cm ) 2
- Description Alpha Beta i Test Loop Pipe Exterior N/A 590 i
i 4-Transferable Alpha and Beta Radioactivity (dpm/100 cm')
Description Alpha Beta i l
Test loop Pipe Exterior MDA MDA
- . Gamma Exposure Rates Average gamma exposure rates for the test loop tank exterior and interior, piping exterior, and ;
i open ended pipe range from 6 nR/hr to 12 R/hr.
Isotopic Analysis of Smear Samples ;
, Gas t:st loop smears were analyzed to identify the radionuclides present and the relative abundance of each. No activity was detected. {
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1 I Chemistry Tr M Isop hping:
i Direct surveys for alpha and beta radioactivity indicated no elevated levels of contamination.
Smear samples analysis results were less than MDA.
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i j Test Imop Dump Tanks and Piping:
Direct Alpha and Beta Radioactivity (dpm/100 cm')
l Description Alpha Beta Dump Tanks and Piping MDA 71,000 Exterior :
Dump Tank Interior MDA MDA Transferable Alpha and Beta Radioactivity (dpm/100 cm')
Description Alpha Beta Exterior of each tank MDA 370 1
3-16 REVISION 0 l
PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY Gamma Exposure Rates A survey of the tanks and piping was conducted and identified an average gamma exposure rate of 0.5 mR/hr and an average beta exposure rate of 2 mrad /hr.
Isotopic Analysis of Miscellaneous Sample.s and Material Scrapings Metal scrapings / shavings md samples of miscellaneous materials removed from the west pit were analyzed to identify the radionuclides present and the relative abundance of each. In all cases, the prunary gamma emitting radionuclides included Cs-137 and Co-60 Test Loop Primary Piping:
Direct Alptja and Beta Radioactivity (dpm/100 cm 2)
Description Alpha Beta Open-Ended Pipe - Return N/A 4,200 Open-Ended Pipe - Supply N/A 11,000 Transferable Alpha and Beta Radioactivity (dpm/100 cm2 )
Description Alpha Beta Open-Ended Pipe MDA 500 Outside of Pipe MDA 42 i Gamma F.xposure Rates Average gamma exposure rates for the open-erxled supply and retum piping range from 68 R/hr l to 78 R/hr.
Isotopic Analysis of Smear Samples Test loop pnmary piping smears were analyzed to identify the radionuclides present and the relative abundance of each. The primary radionuclide identified is Co-60.
Heating and Ventilation Ductwork:
Direct Alpha and Beta Radioactivity (dpm/100 cm 2)
Description Alpha Beta i
Exhaust Ducts 86 980 j 4
Supply Ducts MDA 450 3-17 REVISION 0
PROTECTION OF GCCUPATIONAL AND PUBLIC IIEALTH AND SAFETl' Transferable Alpha and Beta Radioactivity (dpm/100 cm 2)
Description Alpha Beta Exhaust Ducts MDA MDA
( Supply Ducts MDA MDA Isotopic Analysis of Smear Samples Large area wipes were obtained and composited for isotopic analysis. The primary radionuclides identified are Co-60 and Cs-137.
Low-level Radioactive Liquid Drain:
Direct Alpha and Beta Radioactivity (dpm/100 cm 2)
Description Alpha Beta Water Trap Drain N/A 2,600 External Overhead Cylinder N/A 290 Transferable Alpha and Beta Radioactivity (dpm/100 cm 2)
Description Alpha Beta Water Trap Drain MDA MDA External Overhead Cylinder MDA MDA
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Isotopic Analysis of Smear Samples Metal, water and sediment samples were collected for isotopic analysis. The pnmary radionuclides identified include Co-60, Cs-137, and Ag-108m.
Process Vent:
Direct Alpha and Beta Radioactivity (dpm/100 cm2 ;
Description Alpha Beta N Vent Interior N/A 380 Vent Exterior (East) N/A 570
, Vent Exterior (West) N/A 510 l
3-18 REVISION 0
PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY Transferable Alpha and Beta Radioactivity (dpm/100 cm2 , ,
Description Alpha Beta Vent Interior (West) N/A 320 j Vent Exterior N/A 130 Isotopic Analysis of Smear Samples ,
, Polar Crane:
Direct Alpha and Beta Radioactivity (dpm/100 cm')
l Description Alpha Beta l Polar Crane Catwalk 91 N/A l Polar Crane Cab N/A N/A Hot Spots 94 8,000 Transferable Alpha and Beta Radioactivity (dpm/100 cm2 )
i Description Alpha Beta Catwalk MDA 37 1 l
Cab MDA 35 l l
Polar Crane Equipment ,
MDA 38 Gamma Exposure Rates Average gamma exposure rates on the polar crane catwalk are 530 R/hr on contact and 20 R/hr at 1 meter. The general area inside the polar crane cab is 13 R/hr.
Isotopic Analysis of Debris Isotopic analysis of debris from the polar crane identified the primary gamma emitting radionuclides as Co-60 and Cs-137. l 3-19 REVISION 0
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~ PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY 4
j ' 3.2 . RADIATION PROTECTION PROGRAM ,
The responsibility for the site radiation program rests with the Radiation Safety Officer and
!' Radiation Safety Committee as established under NRC License No. SNM-770, and as continued !
under TR-2. !
- i The Waltz Mill radiation protection program will ensure that all radiological activities conducted
' i during the Decommissioning Project comply with regulatory requirements. Radiological hazards
[ will be monitored and evaluated on a routine basis to maivain radiation exposums and the release of radioactive materials to unrestricted areas as far below specified limits as reasonably achievable.
. The radiation protection program will be integrated into all remediation project work activities, and each element of the program will be defined and implemented by approved policies, procedures and guidelines.
The Waltz Mill Radiation Protection (RP) Manual describes the essential elements of the program.
It provides the responsibilities, authorities and qualifications, administrative policies, program objectives and standards to implement the radiation protection program. Included in the RP manual is the commitment of management to incorporate ALARA principles and philosophy into all radiological work' activities. ' This commitment will ensure that the occupational radiation exposures for individual and collective doses and the releases of radioactive effluents are ALARA.
Established Health Physics (HP) procedures will provide guidance for performing specific tasks and methods used to maintain a radiologically safe working environment. HP procedures specify the types of instmmentation and the methods to be employed when perfonning surveys and obtaining samples. Examples of typical HP procedures for surveillance include:
o Radiation, surface and airborne radioactive material surveys o Identification and posting of radiation, high radiation, surface and airborne radioactivity areas o Access controls for radiation, high radiation, surface and airborne radioactivity areas o Hot particle area posting and control o Protective clothing selection, issue, donning and removal o Protective clothing collection, cleaning, survey and reissue o Penonnel radioactivity monitoring and decontamination.
o Radiological protection incidents and reports o Radiation protection surveillance, evaluation and assessment programs The remediation project may involve work activities which are not normally performed during site operations at Waltz Mill. To ensure the current Waltz Mill radiation protection program is adequate to protect the health and safety of worken dunng the remediation project, a review of the current program has been performed. As necessary, program enhancements will be implemented prior to the start of field remediation activities. l
< 'Ihe following sections describe major elements of the program.
3-20 REVISION 0 'l
1 PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY !
s 3.2.1 Eadioloeical Surveillance and Work Area Controls i '
! 3.2.1.1 Radiological Evaluations Radiological area surveys will be perfomied to monitor and record the radiological conditions of l
, the work area and adjacent unrestricted areas, and to meet the site administrative guidelines and the i
requirements 'of 10 CFR 20 (Ref. 2). These surveys will identify and measure direct radiation. l
- levels, airborne radioactivity, and surface radioactivity.
Results will be documented on survey forms or recorded in logs. This information will be available 7
- to personnel entering the radiological area. The information on the survey form may include a '!
sketch or map of the area, contact and general area dose rates, radioactivity levels, identification of l
specific hazards such as hot spots, and the location of radiological boundaries.
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A supervisory review will be performed on all evaluations to ensure that they are adequate to assess the radiological hazards in the area, and that all information is properly recorded. The supervisor i reviewing the survey will ensure that the results are consistent with those anticipated and, if not, !
will determine the reason for the variance. '
i Survey frequencies will be established at the direction of the RSO and will be based on the hazard '
which may be encountered, the potential for changing radiological conditions, and the frequency of l
occupation. Evaluations will be performed to provide positive verification that radioactive ;
materials are being adequately controlled and are not spreading to unrestricted areas. !
l 3.2.1.2 Radiation Work Pennits A Radiation Work Permit (RWP) system will be used for the administrative control of personnel !
entering or working in areas that have radiological hazards present. Work techniques will be :
specified in such a manner that the exposures for all personnel, individually and collectively, are l maintained ALARA. RWPs do not replace work procedures, but act as a supplement to ;
procedures. Radiation work practices will be considered when procedures are developed for work ,
which will take place in a radiologically controlled area. '
l Project RWPs will describe thejob to be performed, define protective clothing and equipment to be l used, and personnel monitoring requirements. RWPs will also specify any special instmetions or precautions pertinent to radiation hazards in the area including listing the radiological hazards present, area dose rates and the presence and intensity of hot spots, loose surface radioactivity, and other hazan.s as appropriate. The HP organization will ensure that radiation, surface radioactivity, and airbome surveys are performed as required to define and document the radiological conditions for eachjob.
RWPs for jobs with low dose and minimal hazards will be approved at the HP technician or HP supervisory level while RWPs for jobs with potentially high dose or significant radiological hazards will be approved by the RSO. Also, collective RWPs for the completion of specific work evolutions with high estimated collective or individual exposures or in high hazard work locations 3-21 REVISION 0 )
PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFET)'
may require review and approval by the Waltz Mill Radiation Safety Committee, in addition to the RSO. Examples of topics covered by implementing procedures for the Radiation Work Permits are:
o Requirements, classifications and scope for RWPs o Initiating, preparing and using RWPs o Extending expiration dates of an RWP o Terminating RWPs 3.2.2 Access Control Areas at Waltz Mill which present a radiological hazard will be posted in such a manner that personnel are made aware of the presence and extent of the hazards in the area. Areas will be posted and access controlled based on the hazard evaluation and will be in accordance with 10 CFR 20 requirements. Access restrictions and entry requirements for areas will be based on the degree of hazard present.
3.2.3 Facilities and Eauipment Sufficient facilhies, equipment, and instmmentation will be available to permit the radiation protection staff to function effectively. The facilities, and types and quantities of instruments provided will be adequate to meet activity needs for the duration of the project. Radiation protection facilities could include the following functions or work areas:
o Sample analysis o Bioassay o Dosimetry issue, storage and calibration o Instmment issue, storage and calibration .
o Access and egress control l o Protective equipment cleaning, maintenance, storage and issue 1
o Personnel change areas j
o Personnel, equipment and materials decontamination area (s) j Radiation protection equipment will include sample counting equipment, portable survey instmments, dosimetry and dosimetry processing equipment, protective equipment, and consumables such as smears and decontamination supplies. A nearby whole body counter is I available under a continuing service contract.
Areas will be provided for the storage, repair, calibration, and issue of the project instmmentation. 1 The operation, repair and calibration of instruments will be performed according to ANSI l standards and manufacturers' recommendations as detailed in procedures or manufacturers' mstmetions. These procedures describe the proper techniques and the limitations for the specific piece of equipment. Calibrations will be appropriate for the anticipated radiation fields and will be traceable to the National Institute of Standards and Technology (NIST).
3-22 REVISION 0
PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTII AND SAFETY 3.2.4 Exposure Contml Exposure control includes both the monitoring and regulation of radiation exposure. Personnel monitoring devices (dosimetry) will be required for all personnel meeting the exposure conditions specified in 10 CFR 20 and in the administrative radiation protection procedures.
Administrative exposure limits will be used to ensure that personnel do not exceed the exposure R' nits specified in 10 CFR 20. The administrative limits will also serve as a management tool to ensure that individual and collective doses are maintained ALARA. Administrative exposure limits have been established in such a manner that increasing exposure levels require increasing levels of mavagement approval.
3.2.4.1 Extemal Whole Body Monitorina Extemal radiation monitoring will be accomplished through the use of primary and secondary dosimeters, such as thermoluminescent dosimeters (TLDs), self-reading dosimeters (SRDs), and electronic dosimeters. The official record of accumulated extemal exposure will normally be obtained from the TLD pnmary dosimeter. Secondary dosimetry will be used as a back-up to the pnmary and as a means for tracking exposure between processing periods.
Persormel are required to wear external radiation monitoring devices whenever work assignments require access to radiologically controlled areas. Primary and secondary dosimeters will be worn on the tmnk of the body between the neck and waist in close proximity to each other unless the RWP specifies otherwise.
Multiple whole body monitoring may be required when work is performed in a non-uniform radiation field and when the portion of the body receiving the highest exposure is not easily determined or is subject to change. RWPs will be used to specify multiple dosimetry requirements (location and position) for personnel working in these areas.
Prunary dosimeters will be processed by a facility which has current accreditation from the National Voluntary Laboratory Accreditation Program (NVLAP). Secondary dosimeters will be calibrated on a semi-annual frequency using NIST traceable radiation fields. ,
Dose information from other sources may replace or supplement primary dosimeter results. Such action may be necessary if the pnmary dosimeter results are unavailable due to loss or damage or if the results are suspect. In these cases, the actions taken and the justification for such actions will be documented and approved by the RSO according to procedures for evaluating exposure when dosimetry is lost, damaged or contaminated.
3.2.4.2 Special Monitorine Extremity monitoring devices will be worn when exposure conditions warrant their use. Specific criteria to be used in determining the need for extremity monitoring and for determining the extremity dose will be identified in procedures or provided on the RWP.
3-23 REVISION 0 i
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l
- _ PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY t
i i j 3.2.4.3 Skin Monitorine i
Monitoring of the skin of the whole body will normally be accomplished with whole body i dosimetry. Instructions regarding the proper method for wearing the dosimeter so that the skin i l dose will be properly measured will be provided in procedures, RWP and/or personnel trainmg.
! Guidarce may also be provided in radiation protection procedures and specified on RWPs for use j
of protective clothing (or other tools for reducing beta, x-ray, and/or low energy gamma radiation) to reduce skin exposure.
l I
- j. When it is suspected that the whole body dosimeter does not provide proper measurements of the i
' skin dose, calculations will be performed according to applica')le radiation protection procedures or using accepted industry computational models.
) 3.2.4.4 Internal Exposures i
j Internal radiation exposure will be minimized by establishing exposure limits and admmistrative i exposure controls, identifying and controlling sources or potential sources of airborne radioactivity,
- - and through the use of engineering controls. Respirators may be used when engineering controls ,
may not adequately protect the worker.
[ l l Breathing zone air sampling, and/or bioassay measurements will be used to determine intakes of radionuclides. Breathing zone air sampling will be the principle means for determining the amount ofintake of radioactive material.
l The bioassay program will provide a quality control check of the success of the program in j tairumizing internal radioactivity of personnel. Bioassay may include whole body counting (in- '
l vivo) and/or analysis of excreta (in-vitro). Bioassay results will be used to evaluate potential j intakes, estimate the magnitude of intakes, and provide data necessary to assess the committed
! effective dose equivalent.
1 Procedures will include criteria for the perfonnance of bioassay, as well as methods for data j analyses, interpretation, and dose assessment. The methods and techniques prescribed by these
- procedures will follow the applicable guidance documents (Regulatory Guides 8.9 and 8.26) (Refs.
f 3 and 4),
3.2.5 Respiratory Protection Program The respiratory protection program will be established in accordance with 10 CFR 20 Subpart H (Ref. 2) and 29 CFR 1910.134 (Ref. 5), and will be based on the guidance provided in ANSI Z88.2 (Ref. 6), NRC Regulatory Guide 8.15 (Ref. 7), and NUREG-0041 (Ref. 8).
Elements of the respiratory protection program include:
o Respirator selection and use o Training programs 3-24 REVISION 0
1 PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY o Medical evaluations o Fit testing o Respiratory protection equipment maintenance and issue records i o' Air quality standards for supplied breathing air systems of Bioassay i
. 3.2.5.1 Respirator User Oualification Respirator users will be . screened and/or examined to establish physical and psychological capabilities necessary to perform tasks using a respirator. Personnel will be trained before using :
respiratory protection devices. Personnel will be fit tested before the first use of respirators requiring a face piece-to-face seal and on an annual basis.
3.2.5.2 Respiratory Protection Eauipment Description and Selection i
4 The requirement for and selection of respiratory protection equipment for radiological purposes will normally be detennined and approved as part of the RWP process. Project supervisors will be responsible for worker compliance with RWP requirements. Only National Institute of Safety and 1 Occupational Health (NIOSH) certified respiratory protection equipment will be used. Routine and emergency issue of respirators will be performed according to applicable procedures. Breathing air may be supplied to respirators from compressed breathing air cylinders, air compressors, or the plant breathing air system. All sourtes of supplied air will meet the requirements for Grade D or better breathing air. l 3.2.5.3 Eauipment Inspection and Maintenance Requirements and techniques for inspection and maintenance of respiratory protection equipment will be performed according to manufacturers' and regulatory requirements. Respirators will be cleaned, sanitized, inspected and maintained according to approved procedures. Respirator repairs will be performed by qualified personnel with parts designed for the respirator Respirators ready for use will be stored to protect against dust, sunlight, heat, extreme cold, excessive moisture, and damaging chemicals.
3.2.6 Radioactive Materials Control Procram Radioactive material control will be implemented through procedures. The radioactive materials control program will be effective in preventing the spread of radioactive materials.
3.2.6.1 Radioactive Material Storage Radioactive material will be stored in specially designated restricted areas. These areas may contain reusable equipment and tools, waste awaiting processing, wastes or other materials prepared for shipment, or equipment and tools awaiting decontmination or reuse. Procedures describe posting requirements, access controls, survey requirements and controls placed on the movement of equipment / materials to and from the storage area.
1 3-25 REVISION 0
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PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY l 1
i 1 3.2.6.2 Contamination Control Program '1 l
The contamination control program is designed to reduce and minimize contaminated areas, tools, l and components, prevent the spread of radioactivity and maintain releases of radioactive materials
- ALARA. Major components of this program include the identification, posting and control of contaminated areas, decontamination of tools, equipment and areas, engineering controls, and l monitoring. Procedures provide guidance for identifying the extent of contaminated areas, j- reducing contaminated areas, release criteria, and decontamination of components, tools, equipment l ' and material.
3.2.6.3Bvoroduct Material Control Procedures specify the requirements for the overall control and accountability of byproduct l materials including material receipt, leak testing, accountability, safe handling, and disposal.
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! 3.2.7 Ensurma that Occupational Radiation Exoosures are As Irw As Reasonably Achievable a
All remediation activities will be planned and conducted in accordance with the ALARA policy'and j comprehensive safety programs. Westinghouse places the highest priority on conducting the Waltz l Mill remediation project safely and maintaining exposures to ionizing radiation ALARA.
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The primary objective of the ALARA program is to muumize exposures of workers, visitors and 4 the general public to ionizing radiation to the maximum extent practicable, taking social, technical and economic factors into consideration. Remediation activities will be conducted in a manner that i ensures the health and safety of all employees,' contractors, and the general public. Westinghouse
- shall ensure that radiation exposures to workers and the public, as well as releases of radioactive j . material to the environment, are maintained below regulatory limits and that deliberate effons are
- taken to funher reduce exposures and releases in accordance with procedures that seek to make any j such exposums or releases as low as reasonably achievable, ,
The ALARA plan consists of several essential elements which will be incorporated into the 4 remediation project and have the full suppon of management. These elements include:
d
[ 3.2.7.1 Management Commitment i
- Management will provide full suppon and commitment to reducing individual and collective j exposures and ensuring appropriate controls to minimize the potential for release of radioactive material to the environment. All Westinghouse and remediation contractor project management will be held responsible for strictly adhering to the ALARA policy.
I. 'All project personnel will be made aware of management's commitment and instructed.on their i responsibility . to execute project activities in accordance with the ALARA policy. This
. commitment will be regularly affirmed through training programs.
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i 3-26 REVISION 0 2
- PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY
- 3.2.7.2 Radiation Safety Committee ne Radiation Safety Committee's main responsibility is to review and panicipate in the
. License No. SNM-770. De committee will also ensure activities are conducted under the tenns &
- conditions of the TR-2 license.
i i 3.2.7.3 Radiological Perfonnance Goals
- j. Radiological perfonnance goals will be established for the remediation project. Performance goals a
will be reviewed to ensure that they are challenging, yet achievable and set by the Radiation Safety 7
Committee.
- Remediation project performance goals may include:
o- Collective exposure for the remediation project
- o Collective exposure for remediation project organizations o Maximum annualindividual exposure
.o Number and type of project personnel contamination events e Number of radiological occurmaces 3.2.7.4 Plans and Procedures The organization, responsibilities, and method of operation of the ALARA program will be addressed in supporting procedures. In addition to the ALARA policy statement, procedures and/or guidance documents will include:
o Responsibility for and performance of ALARA job reviews o Radiation Safety Committee charter o Radiation Work Permits 3.2.7.5 Radiological Work Plannma
- Waltz Mill remediation activities inside the restricted area (radiological work) will be conducted acconting to radiation protection procedures to control workers' exposure and the spread of radioactive material. Work involving significant radiation exposure will be planned as far in advance as practical. During planning, unnecessary work steps will.be deleted, radiation and
- radioactivity levels in the work area will be detennined, and collective exposure estimated. The following considerations may be factored into all work plans:
o Determine needed tools, parts, equipment, etc. before the work begins and stage them to minimize delays o Coordinate effons of different groups, such as decontamination, constmetion, radiation protection, so work can proceed in a systematic and efficient manner 3-27 REVISION 0
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1 PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY 1
o Minimize the number of workers assigned to a particularjob I i o Coordinate work by area so that work, such as scaffold and shielding installation and
! removal, is not duplicated for multiple tasks to be performed in the same hrea i
e Perform as much work outside of radiation areas as possible
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i o Remove or shield sources of radiation to allow work to proceed in lower gamma exposure !
! rate areas o
Identify required procedures and facilitate review and concurrence before work may 3 proceed
! o Identify special tools (including robotics, remote handling equipment, video monitors, etc.) ;
] and temporary services (including auxiliary lighting, power, communications) and ensure i availability and operational status at the work location :
4 :
L are effective tools for communicating radiological information; providing instruction to workers f '
regarding radiological work requirements; defining radiation protection practices, instmments, and equipment to be employed; and special requirements, such as specific procedure requirements, j engineering controls, containments, shielding, etc. For ALARA pmposes, a preliminary estimate
! of time and exposure' for the activity and any special ALARA controls may be provided, as ;
) appropriate. The RWP will be reviewed by the supervisor of the organization responsible for the 1
- work and, if necessary, reviewed by the Radiation Safety Committee or designated HP personnel, as appropriate, before beginning work activities. 1 A fonnal ALARA job review will be conducted for work that has the potential to exceed radiological thresholds established by the Radiation Safety Committee and incorporated m l
} implementing procedures. I i
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4 b l 1 1 3 3-28 REVISION 0 4
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PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND S.4FETY 3.3 RADIOACTIVE WASTE MANAGEMENT This section addresses the technologies, equipment, and procedures to be implemented for the ,
management of radioactive waste during the project. These technical approaches are based upon experience and address facets of plannmg, decontamination, packaging, storage, transportation, i
volume reduction or beneficial reuse, and final disposition of the waste materials, while minimizing a secondary wastes and radiation exposure.
In developing the radioactive waste management program, the following elements will be 4
considered:
o Location and availability of disposal facilities o Potential for off-site release during remediation operations o Preventing contamination of uncontaminated areas o Use of existing facilities to support the waste packaging operations o Methods of approach related to waste type, waste class, and impact on safety i o Cost effectiveness o Logical approach to remediation operations o Ensurmg that the occupational exposures are maintained ALARA
." o Mininuzmg the impact on the health and safety of the general public
! o Maintaining flexibility for waste management to allow for unexpected wastes and changes in available technology o Mininuzation of radioactive waste o Quality Control On-site packaging or processing of radioactive waste prior to transponation will be performed in areas designated for these activities. Except for lead shielding, no sources of mixed waste have been identified. No chemicals or other substances will be used during remediation operations that may become hazardous wastes or result in mixed waste. To reduce or avoid the generation of mixed wastes, poject management will control the use of any chemical or other substance that could become a mixed waste concern.
If hazardous materials containing radioactive material are identified during remediation, they will be classified and stored on-site until declassified or approved for disposition. These materials will be managed according to Federal, state, local and site pennitting requirements to the extent it is not inconsistent with NRC handling, storage and transportation regulations.
3-29 REVISION 0
_ . _ . . _ . ~ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - . _ _ . _ _ . - _ . _ _ ,
l 1 q PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY i
3.3.1 Radioactive Waste Processine i 3.3.1.1 On-Site Radioactive Waste Volume Minimintion i,
i Minimization of the quantity of radioactive waste requiring disposal is a high priority during the 1 j project. Project management will incorporate the radioactive waste volume minimization pract;m
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4 into work procedures. - The following elements will be included as appropriate:
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' Radiation worker training will identify policies and practices to prevent the unnecessary generation of mixed or radioactive wastes.
! o '
Unnemy generation of radioactive and mixed wastes will be minimized by controlling chemicals brought on-site, and preventing unnecessary packaging, tools and equipment i from entering radiologically controlled areas. -i o
Some materials will be reused dunng the remediation pmject. This typically includes !
[
4 contaminated tools, equipment, and clothing (after laundering). !
o The volume of radioactive waste will be minimind by decontaminating areas: and ,
equipment where practical, and by segregating waste as radioactive and non-radioactive, ;
- where practical.
}- o Decontamination activities will be planned to minimize the generation of secondary waste i
- volumes as a result of decontamination processes. l 4
o Bulky material may be dismantled or cut up to reduce volume. Metal materials, !
s incinerables and compactible waste may be sent off-site for processing. )
- o Waste containers for direct burial will be packaged so that void space is mmunized. Space i around large bulky objects will be filled with small items and debris, if conducive to ALARA. I 3.3.1.2Off-Site Shioments of Radioactive Materials for Further Processing l i
The project will result in the accumulation of significant volumes of contammated and/or activated
- material anxi debris. To minimize the volume of material for disposal, licensed waste processors, !
! capable of reducing this volume to the maximum extent practical, will be utilized when cost !
effective. This will be determined through cost benefit analyses performed for each waste stream '!
j and/or waste category. j i
3.3.1.3 Liauid Waste Processing System
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1 1 i- ' Contaminated water may be generated as a result of draining, decontamination, and cutting [
- processes. The contaminated liquids will be processed either in the existing liquid waste processing l system located in the basement of the Facilities Operations Building or in a temporary treatment !
system (e.g., ion exchange and filtration system), solidified or sent to a licensed waste processor.
- All liquid radioactive waste will be processed according to approved procedures for waste collection i
and discharge. Liquids released from site are monitored and controlled to ensure all releases of
' radioactivity to the environment are as low as is reasonably achievable and that the releases meet I j established admmistrative controls, and regulatory criteria.
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, 3-30 REVISION 0 !
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PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY 3.3.1.4 Local Ventilation
) Local HEPA filtration systems will be used when activities could result in the generation of ,
! significant airborne radioactive paniculate activity. If the filtered air is exhausted directly to an '
unrestricted area, appropriate air sampling will be performed at the point of discharge, d
3.3.2 Radioactive Waste Disoosal 3.3.2.1 Waste Classification i
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Proper classification of waste for disposal will be conducted using procedures which implement the !
! requirements of federal regulations and disposal site criteria. Procedures will ensure that a realistic
! representation of the distribution of radionuclides in waste is known and that waste classification is '
performed in a consistent manner. Any of the following basic methods, used individually or in -
i combination, will be used to achieve this goal: materials accountability (including process
' knowledge and activation analysis), classification by source, gross radioactivity measurements, and i measurement of specific radionuclides.
Waste characterization will be pedormed on samples from each area (building), room or process.
. Individual waste stream designations will be established for areas and processes that have dissimilar i j radionuclide distributions and physical properties (dry active waste, liquids, sludges, etc.) :,
e l' Appropriate instrumentation will be used to determine the type and quantity of radioactive material !
in each waste stream. Samples for radionuclide distribution will be obtained from wastes, !
{ whenever practicable. The curie content of each package can be calculated when radionuclide l concentrations are determined, or by using a dose rate to curie conversion factor. Characterization i
- j. will be performed by monitoring and/or sampling before packaging, and the activity of each j radionuclide present in the mixture will then be used to estimate the activity in the fmal package. j
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j Radioactive waste will be classified cs A, B, C or greater than Class C according to 10 CFR 61.
d Table 3-1 contains an estimate of the radioactive waste anticipated to be generated durmg WTR !
l decommissioning. This Table includes the total volume anticipated for one piece reactor vessel,
- ~ biological shield, and vessel internal removal (2050 ft'), the total volume for multiple piece reactor -
i vessel, biological shield, and vessel intemal removal (2584 ft'), and those areas described in Section !
l 2.7. It is anticipated that the majority of this waste will be Class A. !
1 1' 3.3.2.2 Waste Packaging. Transfer and Storage i
- j. Waste will be packaged at the point of generation or at a designated location on site. Packaging ;
j may include approved disposal containers. Other forms of packaging may include radioactivity ,
- - control measures, such as bagging and/or wrapping, to facilitate safe transportation to other !
locations within the site for further processing. Heavily contaminated items may require further !
I radioactivity control measures, such as the application of a fixative, in addition to bagging or
[ wrapping.
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J 3-31 REVISION 0
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PROTECTION OF OCCUPA TIONAL AND PUBIJC HEALTH AND SAFETY After packaging, the waste will be tmnsported to an on-site staging area and prepared for shipment. If necessary, the waste container will be placed in a storage area. Greater than Class C waste, if any, will be stored until it can be transported to a facility licensed to accept greater than Class C waste.
1 Waste storage facilities planned for use during remediation activities include:
o Trailers and sea / land containers may be stored and used on-site to temporarily house dry and solid low level waste o
Selected yard areas may be used for short term storage of packaged waste staged for transpon ;
o WTR Trucklock, or other designated onsite facility for extended interim storage o Temporary storage areas for building mbble and soil Once all waste is removed from the storage locations, the areas will be surveyed and l decontaminated, if necessary, and temporary structures removed.
- 3.3.2.3 Waste Transportation l Before waste is shipped from Waltz Mill, each package will be inspected to ensure it meets all l applicable design and/or certification requirements and the container is not damaged or impaired.
- Most shipments are expected to be low specific activity (LSA) and will be shipped in exclusive-use vehicles. Radioactive material and waste will be transported by truck or rail, depending on the volume of material, vehicle availability and packaging requirements. In some cases, approved .
- shielded casks will be employed due to radiation levels or limits for quantities of radioactivity in a i package.
Some of the relevant regulatory requirements are discussed below. l
) o DOT Regulatory Requirements - All radioactive matenal/ waste shipments shall be l
! performed in accordance with DOT and other applicable federal regulations, as well as t
l burial site requirements.
o Documentation - Radioactive waste shall only be shipped to a licensed waste processing or i disposal facility. Where applicable, a state user's permit may be required. All required documents shall be complete and legible and shall meet the requirements in 49 CFR,10 l CFR 20, and the receiving facility license requirements.
! e Shipping Routes - ne actual routing of shipments may vary with weather and highway
, conditions. Additionally, local and state restrictions penaining to radioactive material
- transport may affect some route selections, particularly in congested metropolitan areas.
i he carrier is responsible for selecting the appropriate route, which must conform to
- applicable federal, state, and local shipping regulations and requirements.
4 o Burial Site or Waste Processor Acceptance Criteria - All packaging, waste form and transportation methods will be in accordance with the criteria for the intended burial or processing facility prior to shipment. The most current, applicable regulations and specific facility terms and license conditions will be used when the shipments are made.
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3-32 REVISION 0
PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY i
o Quality Control - The quality control program for waste packaging and shipping will j
provide assurance and verification of compliance with radioactive waste shipping j regulations. The remediation contractor will have the appropriate DOT certification for all
- waste packaging.
! 3.3.3 Disposal of Non-Radioactive Waste i
I Non-radioactive wastes will be disposed of by release to appropriate disposal facilities such as l landfills, scrap yards and scrap recovery facilities. Materials that are inappropriate for surface j surveys will be sampled and appropriately analyzed. Materials found to be non-contaminated will j be disposed of as non-radioactive waste.
3.3.4 Release of Material for Unrestricted Use -
Surface contamination surveys will be conducted for both removable and fixed contamination l before potentially contaminated equipment is released from restricted to unrestricted areas. Release :
of equipment and packages from the Waltz Mill site to unrestricted areas shall be in accordance with Reference 9. 'Ihe survey methodology used will be sufficient to detect the levels specified in Table 1 of Reference 9. ,
j 3.3.5 Hazardous Waste l All hazardous waste generated as a result of this activity, will be handled, packaged, transported, and disposed of in compliance with Pennsylvania Code, Title 25, Environmental Resources, Chapters 260--270, Hazardous Waste Regulations. Activity will be identified with ;
current EPA ID Number PAD 074953241 issued to the Waltz Mill Site.
3.3.6 Mixed Waste All hazardous waste generated as a result of this activity, will be handled, packaged, transported, and disposed of in compliance with the consent order and agreement between Westinghouse Electric Corporation and the Department of Environmental Protection, dated August 28,1996. At all times, Westinghouse will manage radioactive mixed waste in a ;
manner that complies with the Solid Waste Management Act,35 P.D. ss 6018.101 et. seq, and ,
the Department's hazardous waste regulations,25 PA Code ss 260.1 et sea.
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3-33 REVISION 0 i i
PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTII AND SAFETY 3,4 ACCIDENT ANALYSIS The risk of accidents resulting in a significant radiological release and off-site dose during decommissioning activities is very low. Since decommissioning activities will not involve any irradiated fuel, only non-fuel related accident scenarios are evaluated in this section. The focus of these decommissioning accident analyses is on public health and safety.
The following postulated accident scenarios have been analyzed considering the radionuclide levels and isotopic composition of components to be processed, and the anticipated decommissioning activities:
- 1. Dropping of contaminated concrete block / rubble
- 2. Fire / Explosion
- 3. Canal Sediment Criticality and Handling
- 4. Rupture of a HEPA vacuum bag The components with the highest radionuclide levels were used in the accident analyses.
Therefore, accidents that were analyzed bound the radiological consequences from other postulated accident scenarios. In evaluating the postulated accidents, conservative assumptions were made when data or knowledge to suppon more realistic analyses were lacking.
Conservatism in this context is defined to mean that the' radiological consequences from the postulated accidents will be overestimated rather than underestimated.
The activity concentrations of the various components used in the following accident analyses were derived from the characterization data summarized in Section 3.1 and described in l Reference 1.
3.4.1 _Ajsumptions The following are the major assumptions used in the following accident analyses:
- 1. All irradiated fuel is removed from the Waltz Mill site.
- 2. Although local HEPA filtration may be used for decommissioning activities, no filtration of radiological effluent is assumed for the accident analyses.
- 3. A worst case atmospheric dispersion factor of 3.53 E-02 sec/m;. This atmospheric dispersion factor was calculated using the guidelines presented in Regulatory Guide 1.145 (Ref.10) and is based on the assumption that the dose receptor is located 100 meters from the radioactivity release point, the wind speed is one mile per hour, and the atmospheric stability condition is extremely stable (Pasquill's turbulence type G).
A distance of 100 meters from the release location is within the 850 acre Westinghouse site property and the Waltz Mill average annual wind speed is greater than one mile per hour.
- 4. All releases to the environment are assumed to be ground level releases.
- 5. A standard breathing rate of 3.33 E-04 m /sec (Ref.11).
3-34 REVISION 0
PROTECTION OF OCCUPATIONAL MD PUBLIC HEALTH AND SAFETY
- 6. The worst case effective committed dose equivalent dose conversion factors for inhalation (Ref. I1).
- 7. Any radiological contaminants as a result of activation of the metallic reactor vessel would remain intact and are not available for release in any of the postulated accidents, except for a release due to HEPA vacuum filter bag rupture.
3.4.2 Dropping of a Contaminated Concrete Block / Rubble Accident Decommissioning activities at the WTR will result in the accumulation of contaminated concrete and the handling of contaminated concrete. This contaminated concrete may be in the form of a concrete block or rubblized concrete. This accident analysis used the worst case radiological activity from samples of the biological shield. This radiological activity concentration was assumed to be homogeneous in all of the concrete block, even though the actual activity levels were next to the reactor vessel. It was assumed that a 50 ton block of concrete is handled. The anticipated capacity of the containment polar crane is 25 tons and the postulated scenario of removing the activated biological shield and activated vessel in its entirety results in approximately 44 tons of concrete (out of the total lift of 148 tons). As this concrete block is being handled with a crane, it is postulated to be dropped and one percent of ;
the activity becomes airborne and is released to the atmosphere. The resultirg dose to an individual located 100 meters down wind of the release is approximately 22 mrem total effective dose equivalent (TEDE). To show the conservatism in this accident analysis, the postulated drop results in 1000 pounds of concrete dust becoming airborne and released to the I atmosphere.
J 3.4.3 Fire / Explosion Accident The majority of the WTR containment is comprised of non-combustible material and fire detection and suppression methods used during any thermal cutting activities make the possibility of a fire or explosion during decommissioning activities low. However, it is postulated that combustible waste (rags, wipes, anti-contamination clothing, etc.) have come in l contact with contaminated surfaces and hold small quantities of radionuclides. This material )
used for decommissioning is placed in a sealand container, postulated to ignite and burn. It is also assumed that 0.1 percent of this entire activity is released to the atmosphere. This release percentage is conservative in that Reference 12 states that 0.015 percent of the activity would be released. The resulting dose to an individual located 100 meters down wind of the release is very small (less than 1 mrem).
It was determined that the consequences of the concrete drop scenario bound those for a postulated local explosion which involves the biological shield concrete or an explosion which would release the accumulated contents of any HEPA filters. It is assumed that a cloud of contaminated concrete, based on uniform radionuclides from the same worst case biological shield concrete samples, is created and carried off site. The drop scenario envelopes the consequences of an explosion due to the conservatism in the drop scenario calculations.
3-35 REVISION 0
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PROTECTION OF OCCUPATIONAL AND PUBUC HEALTH AND SAFETY :
i 3.4.4 Canal Sediment Criticality and Handling Since the canal sediment and canal water are categorized as optional areas to decommission m accordance with the TR-2 Decommissioning Plan (see Section 2.7), the analyses required to assure suberiticality of the canal sediment will be performed prior to remediating the canal.
Additionally, the handling, stabilization, shipment, and disposal of the sediment will be i i
evaluated prior to remediating the canal. :
1 3.4.5 Ruoture of a HEPA Vacuum Bac I For this scenario it is assumed that a HEPA vacuum collection bag ruptures at the time it is full, just prior to change-out. The resulting dose to an individual located 100 meters down !
wind of the release is very small (less than 1 mrem). !
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I 3-36 REVISION 0 i
PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY REFERENCES FOR SECTION 3
- 1. Westinghouse Electric Corporation, Waltz Mill Facility, Characterization Report, Nuclear Materials License TR-2, Test Reactor, Volumes 1 and 2, dated Febmary 9,1994.
- 2. 10 CFR 20, " Standards for Protection Against Radiation."
- 3. NRC Regulatory Guide 8.9, " Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program," dated July 1993.
- 4. NRC Regulatory Guide 8.26, " Application of Bioassay for Fission and Activation Products," dated September 1980.
- 5. 29 CFR 1910, " Occupational Safety and Health Standards."
- 6. ANSI Z88.2, "American National Standard for Respiratory Protection," dated August 1992.
- 7. NRC Regulatory Guide 8.15, " Acceptable Programs for Respiratory Protection," dated October 1976. >
- 8. NRC NUREG-0041, " Manual of Respiratory Protection Against Airborne Radioactive Materials," dated October 1976.
- 9. NRC document, " Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Temunation of Licenses for Byproduct, Source, or Special Nuclear Material," dated May 1987.
- 10. NRC Regulatory Guide 1.145, " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," November 1982.
- 11. EPA Federal Guidance Repon No.11, " Limiting Values of Radionuclide Intake And Air Concentration and Dose Conversion Factors For Inhalation, Submersion, and Ingestion,"
September 1988. ;
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- 12. NRC NUREG/CR-1756, " Technology, Safety and costs of Decommissioning l Reference Nuclear Research and Test Reactors," March 1982. l l
3-37 REVISION 0
PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY Table 3-1 ESTIMATE OF RADIOACTIVE WASTE VOLUME COMPONENT VOLUME (ft 3)
Reactor Vessel Body 210 Access Plug & Flange 97 Upper Vessel Internals 265 Lower Vessel Flange Area 85 Internal Shields & Lattice 742 Biolegical Shield 1.185 REACTOR VESSEL, BIOLOGICAL SHIELD, and INTERNALS SUBTOTAL
- 2,584 5
Transfer Canal 110 Primary Coolant Piping 4,337 Thimble Valve Banks 7 Rabbit Pump & Piping 6 Gas Test Loop & Piping 2 Dump Tank & Piping 1,703 Test Loop Primary Pipe 4 Experimental Cooling Water Pipe 6 Low Lesel Radiological Drain 156 Process Vent 26 Electrical Conduit & Boxes 432 Plant & Instrumer' Air 9 Steam & Condensate Piping & Valves 15 Polar Crane 155 Sub-Pile Room 246 Beam Port Platform (elevation 37') 119 Head Stand Platform 12 Top Platform (elevation 61') 10 OPTIONAL AREAS SUBTOTAL 7,355 qTOT L 9,939 (1) The total volume for one piece reactor vessel, biological shield, and vessel internal removal (Option 1) is 2050 ft'. The total volume for multiple piece reactor vessel, biological shield, and vessel internal removal (Option 2) is 2584 ft'.
NOTE: It is anticipated that the majority of the waste will be Class A.
3-38 REVISION 0
PROPOSED FINAL SURVEY ;
SECTION 4 PROPOSED FINAL SURVEY l l
The WTR decommissioning activities will result in the removal of the reactor biological shield, ;
vessel and internal components. Also, decontammation and dismantlement activities of other !
stmetures and equipment associated with TR-2 may be performed under the provisions of tie i WTR Decommissioning Plan. After removal of the reactor vessel internal contents, the reactor i
vessel, ar.d the biological shield, all remaining residual radioactivity and WTR facilities will be !
transferred to the SNM-770 License, where it will be addressed by the SNM-770 Remediation Plan. Upon completion, no materials covered by the 10 CFR 50 license will exist.
The method for determining that the WTR facility has met the decommissioning objectives and I prerequisites for license termmation will be an independent verification that the reactor vessel internal contents, the reactor vessel, and the biological shield have been removed. This independent verification will be performed and documented in accordance with the Project l Quality Plan.
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t 4-1 REV.O
FUNDING SECTION 5 FUNDING Westinghouse has established one Financial Assurance Mechanism that er. compasses all of the Westinghouse facilities that hold NRC licenses. The Financial Assurance Mechanism established by this approach meets all the requirements of the NRC's decommissioning financial assurance regulations contained in 10 CFR 50. Appropriate updates have been submitted to the NRC to maintain adequate levels of financial assurance.
In March of 1997, Westinghouse submitted to the NRC a revision to the Finarrial Assurance Mechanism for Decommissioning its NRC licensed facilities (Ref.1). This submittal was reviewed by the NRC staff and found to be in compliance with the regulations (Ref. 2). Future updates will be made as appropriate.
5-1 REV.O
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FUNDING REFERENCES FOR SECTION 5
- 1. !
Westinghouse letter, Nardi to NRC, dated March 6,1997;
Subject:
" Revised Financial !
Assurance Mechanism for Decommissioning."
'2 .
NRC letter, Hickey to Nardi (Westinghouse), dated April 23,1997:
Subject:
" Response to Revised Financial Assurance Mechanism for Decommissioning." .
i 5-2 REV.O
TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS SECTION 6 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS 1
The proposed WTR Technical and Environmental Specifications are included as Appendix A to the TR-2 Decommissioning Plan.
The WTR technical specifications are applicable to all remaining activities at the WTR,
~, including safe storage and decommissioning activities. This is consistent with the requirements of 10 CFR 50.36.(c )(6), which provides for case-by-case development of technical specifications for non-power reactors that are not authorized to operate.
Limiting conditions for operation and surveillance requirements are identified for 1 conf'mement, for ventilation systems, and for radiation and effluent monitors. These provisions provide reasonable assurance that radioactive contammation will not be released to the environment in excess of 10 CFR Part 20 limits. Flexibility is provided to permit dismantlement activities and removal of componen and material from the reactor
- building, while maintaining ventilation and admmistrative controls on activities that could i
result in airborne contamination.
The accident analyses in Section 3.4 show that the source terms at the WTR are
} sufficiently small that no equipment features, such as filtered ventilation, are relied upon j to maintain public doses well below regulatory guidance limits. Therefore, the i requirements of the technical specifications are conservative limitations to protect workers and the public from any radioactive material sources discovered during 6 commissioning i; that might substantially exceed expected activity levels.
i The environmental specifications included in Appendix A are the same sampling and monitoring requirements currently being performed for the Waltz Mill site under License No. SNM-770, with the exception that a weekly gaseous air monitoring requirement is added during activities that could produce airoorne contammation in the WTR reactor
- building in excess of 10 CFR 20 limits. The current program periodically monitors air, i water, soil, sediment, and vegetation representative environmental samples. l Admuustrative controls include requirements for independent review and exammation of WTR programs and activities through a Review Committee. Procedures, reports, and i records requirements are also identified, as appropriate, for decommissioning activities.
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l 6-1 REV.O 1
QUALITYASSURANCE PLAN SECTION 7 QUALITY ASSURANCE PLAN i
A Project Quality Plan (PQP) will be developed to incorporate the applicable portions of 10 CFR 50, Appendix B. In addition, the PQP will id::ntify additional procedures and requirements that I are applicable based on government and regulatory requirements, contractual commitments and supplemental quality standards.
i The following is a list of the sections from the Westinghouse. Quality Management System and whether they are applicable to decommissioning activities at the WTR:
l 1.0 Management Responsibilities - Applicable 2.0 Quality Systems - Applicable 3.0 !
Contra.et Review - Applicable 4.0 Design Control- Applicable 5.0 Document and Data Control - Applicable 6.0 Purchasing- Applicable .
7.0 Control of Customer-Supplied Product - Not Applicable 8.0 Product Identification and Traceability - Applicable !
9.0 l ' Process Control - Applicable 10.0 Inspection and Testing - Applicable 11.0 Control ofInspection, Measuring, and Test Equipment - Applicable 12.0 Inspection and Test Status - Applicable 13.0 Control of Nonconforming Product - Applicable 14.0 Corrective and Preventive Action - Applicable 15.0 Handling, Storage, Packaging, Preservation and Delivery - Applicable 16.0 Control of Quality Records - Applicable 17.0 Internal Quality Assessments - Applicable 18.0 Training - Applicable i
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19.0 Servicing-Not Applicable !
20.0 Statistical Techniques - Applicable An extensive quality assurance program will be carried on throughout the TR-2 decommissioning effort to assure that work does not endanger public safety, and to assure the safety of the
. decomnussioning staff.
I 7-1 REV.0
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QUALITY ASSURANCE PLAN i
Quality Assurance efforts during the TR-2 decommissioning period will include the following: '
o performing QA functions for procurement j
- o. qualifying suppliers e auditing all project activities I o
monitoring worker perfonnance for compliance with work procedures o
verifying compliance of radioactive shipments with appropriate procedures and regulations o
performing dimensional, visual, nondestmetive exammations or other required inspection services to assure compliance with work plans o maintaining auditable files on the QA audits o
preparing a final report on overall performance of the TR-2 Decommissioning Project with regard to the QA function i
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7-2 REV.0
ACCESS CONTROL PL4N SECTION 8 ACCESS CONTROL PLAN i
8.1 CURRENT PROVISIONS Access to the Waltz Mill site is currently controlled in accordance with an industrial I security program. The entire facility is surrounded by a security chain link fence and is protected by a security guard force.
Entrances into the WTR contairunent building are locked and access is controlled by the :
RSO. Entrances to certain areas within the containment building are also locked for radiological control purposes to preclude inadvertent entry.
The Access Control Plan described in this section will address controls related to-decommissioning activities in the WTR containment building. Access control !
requirements into radiologically controlled areas are based on 10 CFR 20 requirements l and are aescribed in Section 3.2, Radiation Protection Program.
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l 8-1 REV.O i
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ACCESS CONTROL PLAN 8.2 ACCESS CONTROL PLAN l 8.2.1 WTR Access Control Ornnintion The Waltz Mill Site Operations Manager is responsible for site access control, including:
Gatehouse and vehicle access into the decommissioning area, and i
e Emergency, medical, and fire reporting.
Access control personnel will be properly trained and will demonstrate understanding of decommissioning area access control requirements and responsibilities. Access control
[ personnel will be unarmed and _ . equipped for continuous on-site and- off-site l communications. Local law enforcement authorities should be familiarized with procedures and plant layout, and arrangements will be made to obtain their services, in
- the event they may be required.
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8.2.2 Physical Security Measures
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8.2.2.1 Physical Barriers l
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- Physical barriers will be used to control access to the decommissioning area as follows
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- The security chain link fence that surrounds the entire Waltz Mill facility will be maintained during decommissioning.
A personnel access gatehouse is and will be located at the main plant entry and will normally be occupied by access control personnel.-
e A vehicle access gate is and will be located in the immediate vicinity of the personnel access gatehouse at the main plant entry. Use of this gate will also be controlled by access control personnel.
The decommissioning area is and will be surrounded by a continuous permanent fence to prevent unauthorized access to restricted areas.
Other plant gates associated with the decommissioning area will be kept locked or continuously monitored by access control personnel.
8.2.2.2 Access Authorization
- Access to the WTR decommissioning areas will be controlled and permitted only to those individuals authorized by the Waltz Mill Site Radiation Safety Officer, or an authorized
. representative.
, 8-2 REV.O
ACCESS CONTROL PLAN 1
All persons passing through the gatehouses will be required to demonstrate valid access authorization. To ensure that only authorized individuals are granted access to the decommissioning site restricted area, decommissioning workers will be controlled through positive identification (e.g., picture badges, controlled access lists, or other means).
Visitor access to the decomrmssioning area must also be approved by the Waltz Mill Site Radiation Safety Officer or a designated representative.
Access to the decommissioning site restricted area does not guarantee access to radiologically controlled areas. The radiation protection staff will continue to administer i
the radiologically controlled area access control program. Specific requirements that must be met prior to accessing radiologically controlled areas are identified in Section 3.
8.2.3 Communications Telephone service will be available at the main plant entry access gatehouse to allow local law enforcement authorities and other local emergency services to be contacted. Radio communications should be available for access control personnel in the event it becomes necessary to limit access to the decommissioning area or if it becomes necessary to contact local emergency services.
8.2.4 Procedures Written procedures will be prepared and implemented to provide access control personnel with guidance for routine and abnormal occurrences. These procedures will include:
. Criteria for identifying abnormal conditions within the decommissioning area,
. Access control personnel actions, and
. Required notifications.
The types of routine occurrences to be addressed in procedures include:
Personnel access control e Vehicle access control Communications equipment and routine testing requirements Surveillance /inspaction of decommissioning area physical barriers Recordkeeping requirements 8-3 REV.O
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ACCESS CONTROL PLAN ,
The types of abnormal occurrences to be addressed in procedures include:
i e Fire or explosion e Site evacuation ~
e Site radiological emergencies e Personnel disturbance
. Acts of perceived threat of sabotage i
-e Civil disturbance Suspected or confumed intrusion sabotage attempt e Breached security area barrier
- Unidentified person in security area e Medical emergencies
. Theft of material 8.2.5. Channes to Current Program The Access Control Program for decommissioning does not involve any changes to the current program that may reduce its effectiveness.
i 8.2.6 Access Control Transition Upon completion of decommissioning activities in the WTR reactor building, all access control program requirements will be transferred to the access control program for the l remainder of the Waltz Mill site. i
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APPENDIX A TO WTR DECOMMISSIONING PLAN TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS FOR WESTINGHOUSE TEST REACTOR l
l REVISION 0 l
WESTINGIIOUSE ELECTRIC CORPORATION WALTZ MILL SITE l
USNRC LICENSE NUMBER TR-2 DOCKET 50-22 !
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i TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR ;
1.0 Introduction ,
These Technical Specifications are applicable to activities at the Westinghouse Test Reactor (WTR), ;
Waltz Mill Site, under provisions of NRC License No. TR-2. The WTR was shut down in 1962, and '
has been maintained in a safe storage condition . !nce that time.
j These Technical Specifications apply during the safe storage period, and also during decommissioning activities. Decommissionmg meludes the dismantlement and removal of the !
j reactor vessel internal contents, the reactor vessel, and the biological shield. All residual radioactivity will then be transferred to the materials license for the remainder of the Waltz Mill Site,
' i NRC License No. SNM-770. After completion of decommissioning and transfer of residual ,
radioactivity to the materials license, the TR-2 license will then be terminated.
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l TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR i I t
2.0 Dermitions
' 2.1 Channel. A channel is the combination of sensor, line amplifier, and output devices which are connected for the purpose of measuring the value of a parameter.
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2.2 Channel Test. A channel test is the introduction of a signal into the channel for verification that it is operable. {
i 2.3 Channel Calibration. A channel calibration is an adjustment of the channel such that its [
output corresponds with acceptable accuracy to known values of the parameter which the ;
channel measures. Calibration shall encompass the entire channel, including equipment l
actuation, alarm or trip, and shall be deemed to include a channel test. !
2.4 !
Channel Check. A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison l
of the channel with other independent channels or systems measuring the same variable. -
2.5 Confinement. A closure on the overall facility or volume within the facility which prevents l
the uncontrolled spread of contamination, and controls the movement of air into and out !
through a controlled path.
1 2.6 Operable. Operable means a component or system is capable ofperforming its intended function.
2.7 Operatine. Operating means a component or system is performing its intended function. l t
2.8 Shall. should and may. "Shall" is used to denote a requirement; "should" to denote a recommendation; and "may" to denote permission, neither a requirement nor a j
recommendation. '
2.9 Facility Specific Definitions i
- a. Restricted Activity. An activity inside of the reactor building involving activated or contaminated reactor facility structures, components or systems that could cause airborne !
material in concentrations in excess of the Derived Air Concentrations (DAC) in 10 CFR 20, t Appendix B, Table 1, Column 3. I
- b. Containment Buildinc. The same structure described in the Final Safety Report as the vapor i containment, or reactor containment building. I 1
- c. Reactor Facility. The contaimnent building, ventilation system, the WTR canal, and !
contaminated piping connecting these components. i A-2
TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- IVTR
- d. Containment Device. An engineered barrier that does not necessarily constitute total enclosure, used to prevent the spread of radioactive contamination and airborne radioactivity.
l e.
Unrestricted Area. An area to which access is neither limited nor controlled by the licensee for purposes of protection ofindividuals from exposure to radiation and radioactive materials.
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TECHNICAL AND ENVIRONhfENTAL SPECIFICA TIONS- WTR 3.0 Limitine Conditions for Operation 3.1 Confinement 3.1.1 Applicability This specification applies to the containment building.
3 1.2 Objective The objective of this specification is to define the activities which require confinement, the conditions necessary for maintaining confinement, and the actions to be taken if confinement is not maintained.
3.1.3 Specifications 3.1.3.1 Activities that Require Confinement Restricted Activities require confinement.
3.1.3.2 Limiting Conditions for Maintaining Confinement During restricted activities:
(1) Either the inner or the outer door (s) in each air lock and in the truck lock shall be kept closed except during personnel ingress or egress, or while equipment is being passed through the doorways.
(2) The ventilation system shall be operating.
(3) While containment openings are open for removal ofmaterials or equipment from the containment building, the ventilation exhaust fans shall be operating ) supply fans, if any, shall be turned off), and all other Restricted Activities shall be suspended.
(4) The outer doors in the air lock and truck lock outer doors shall be locked or blocked closed to prevent unauthorized entry except when authorized personnel are inside the containment building or outside with the door in view.
(5) When the ventilation system is not operating, the containment building shall be maintained in a condition such that there are no airflow pathways open directly to A-4
TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- IITR areas external to the containment building except the air lock doorways when the doors are open.
3.1.3.3 Actions To Be Taken If Confinement Is Not Maintained If Confinement is not maintained in accordance with Specification 3.1.3.2:
(1) Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, all Restricted Activities shall be suspended.
(2) Other activities that are NOT Restricted Activities may proceed without Confinement, provided that ventilation exhaust fan (s) are operating.
3.1.4 Bases Historical measurements of airbome radioactivity inside the containment building indicate that the containment building in its present condition does not require a confinement system. Because of the activation and surface contamination levels, dismantlement activities associated with removal of reactor vessel internal contents, the reactor vessel, biological shield, and other Restricted Activities could cause airborne concentrations in excess of DAC. Maintaining confinement during Restricted Activities prevents the uncontrolled spread of contamination during these actisities.
The restrictions and limitations in Specification 3.1.3.2 are necessary to provide assurance that an effective confinement system will be established and maintained.
The one hour action time provided in Specification 3.1.3.3 allows an orderly suspension of activities in the event that the conditions specified for maintaining confinement are not met. If confinement is not maintained, other activities are permitted as long as ventilation exhaust fans are operating to minimize the potential for outward air flow; this provision permits activities that would require an opening in the containment building to remove items that have no potential for creating airborne contamination in excess of 10 CFR 20 limits. For example, the reactor vessel could be removed by suspending Restricted Activities that could create airborne contamination, cutting a suitable opening in the containment building, and maintaining the ventilation system operating, while the reactor vessel is removed from the containment building.
3.2 Ventilation Systems 3.2.1 Applicability This specification applies to ventilation systems used to prevent uncontrolled spread of airbome contamination within the reactor building .
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I TECHNIOL AND ENVIRONMENTAL SPECIFICA TIONS- WTR 1
3.2.2 Objective 1 This specification describes the minimum requirements for operation and installation of ventilation systems.
l 3.2.3 Specification I 3.2.3.1 Activities that Require Ventilation Activities involving physical handling of the reactor vessel, reactor vessel internal contents, biological shield, or any Restricted Activities require that a ventilation system be operating.
3.2.3.2 Limiting Conditions for Ventilation Systems (1) During activities that require ventilation, ventilation system (s) shall be operating to ensure that air flow is from zones oflesser potential for airborne contamination to zones of greater potential for airborne contamination.
(2) Ventilation systems shall be designed to confime radioactive materials and to prevent uncontrolled release of radioactive material. Materials of construction shall be fire-resistant.
(3) Ventilation systems may be of a localized type using temporary containment devices.
(4) Ventilation systems shall discharge through a particulate filter system capable of ensuring that air effluents comply with the requirements of 10 CFR 20, Appendix B, Table 2, Column 1 and 10 CFR 20.1101, for unrestricted areas.
3.2.3.3 Actions To Be Taken If Ventilation System Requirements Are Not Maintained If ventilation system requirements are not maintained, within I hour, suspend all activities within i
the area served by the inoperable ventilation system, that involve physical handling of the reactor I vessel, reactor vessel internal contents, biological shield, and all other restricted activities. I If the exhaust release rates are such that effluent limits may be exceeded, immediately suspend activities causing the release. Implement corrective action to ensure further release is within the limits.
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TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR 3.2.4 Bases The extremely small source term at the WTR is adequately confined by the containment building.
Whenever restricted activities are in progress, an operating ventilation system will minimize the ,
spread of contamination to other areas, and the discharge filter and effluent monitor (see Specification 3.3) provide assurance that efDuent concentrations are within applicable limits. The decommissioning accident analyses do not take credit for filtered ventilation of any accidentally released radioactive materials, so no specific HEPA filter efficiencies are required The only requirement for exhaust filter installations is that air effluents must comply with the requirements of 10 CFR 20.
The restrictions and limitations in Specification 3.2.3.2 are necessary to provide assurance that an effective ventilation system will be established and maintained.
The one hour action time provided in Specification 3.2.3.3 allows an orderly suspension of activities in the event that the conditions specified for maintaining ventilation system requirements are not met. Immediate actions are appropriate to correct effluent releases that could exceed 10 CFR 20 limits.
3.3 Rcdiation and Effluent Monitorine Sysg_n)s 3.3.1 Applicability This specification applies to those devices either permanently installed or portable, used to detect radiation and/or contamination levels, and to monitor effluents released, if any, from the containment building.
3.3.2 Objective To describe the minimum radiological instrument capabilities that must be available for use at the reactor facility to protect workers and to ensure that effluents released from the containment building meet regulatory requirements.
3.3.3 Specifications 3.3.3.1 Activities that Require Monitoring Systems Radiation and effluent monitoring systems are required during physical handling of reactor vessel internal contents, the reactor vessel, biological shield, and during any other Restricted Activities.
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TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR 9
J 3.3.3.2 Limiting Conditions for Monitoring Systems !
The radiological instrumentation capability that must be available for use at the reactor facility is as s follows:
i (1) Airborne Activity Monitors - Both portable and/or stationary effluent, general area, continuous air monitoring devices, and personal air sampling devices shall be used, as necessary, in the containment building, and will be appropriately located to support activities in progress.
, (2) Portable Instrumentation - An adequate number ofinstruments of sufficient accuracy l
and sensitivity shall be available to ensure compliance with the radiation monitoring j j and measuring requirements of 10 CFR Pan 20 including beta-gamma survey meters (up to 1000 R/hr) for radiation dose rates and surface contamination measurement (up i
to 500,000 cpm) and alpha survey meters for surface contamination (up to 50,000 l cpm).
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- (3) Lab Counting Instrumentation / Methods - Gamma spectroscopy, and other standard lab counting methods, including wet chemistry methods. l i l l
. 3.3.3.3 Actions To Be Taken If Required Radiation Monitors Are Not Operable If stationary monitors are inoperable, within I hour, install suitab!: portable instruments, or perform surveys or analyses under direction of the Radiation Survey Officer, as substitutes for any of the monitors in this section.
With no operable radiation monitors, or applicable surveys or analyses, suspend all activities involving physical handling of reactor vessel internal contents, the reactor vessel, biological shield, or any other restricted activity, l
3.3.4 Bases The monitoring systems described in 3.3.3.1 will provide assurance that the radiation levels and the concentration of airborn radioactive material in the working areas are measured during the conduct of restricted activities and all other physical decommissioning activities. l l
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TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR ,
i 3.4 Effluents and Environmental Monitorinc t
3.4.1 Applicability !
This specification applies at all times. '
3.4.2 Objective I k
To ensure that air and liquid effluents released from the TR-2 reactor facility conform to the requirements of 10 CFR 20, and that environmental monitoring is performed to confirm the '
effectiveness of eflhnt controls.
3.4.3 Specifications 3.4.3.1 Limiting Conditions for Air Effluents Radioactive material discharged from the containment building to the atmosphere shall conform to the requirements of 10 CFR 20, Appendix B, Table 2, Column 1 and 10 CFR 20.1101.
3.4.3.2 Limiting Conditions for Liquid Effluents !
Liquid effluents exceeding the effluent concentration limits of Column 2 of Table 2, Appendix B, 10CFR20, shall be processed through the site liquid waste processing system and discharged in accordance with the provisions of License Number SNM-770.
3.4.3.3 Actions If air or liquid effluents are determined to exceed the limitations of Specifications 3.4.3.1 or 3.4.3.2 above, activities that produce those effluents shall be immediately suspended.
l 3.4.4 Bases Effluents produced during dismantlement and other decommissioning activities must continue to meet the radiation protection requirements of 10 CFR 20. The Waltz Mill environmental monitoring program specified in Specification 4.4, and conducted in accordance with license SNM-770 will continue to verify the environmental impacts, if any, of radiological releases from the facility. This j program examines air, water, soil, sediment, and other representative environmental media in the i
- surrounding area.
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TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR
- 4.0 Sun >eillance Reauirements 4.1 Confinement '
! l 4.1.1 Applicability i
This specification applies to the surveillance requirements for the reactor building confinement
} system. I I
l a 4.1.2 Objective To assure that the reactor building is maintained in a condition that provides an effective boundary j for the confinement system.
) 4.1.3 Specifications i (1) At least annually and prior to initiation of any restricted activities, facility records j
shall be reviewed and the containment building shall be visually inspected to determine that there are no pathways open directly to the environment, except the air
- l. lock and truck lock doorways when opened.
T.
- (2) At least once per month, a visual examination shall be performed to determine that the outer air lock and truck lock doors are locked or blocked closed whenever no one
- is inside of the containment building.
- 4.1.4 Bases 4
Compliance with these specifications provides assurance that the containment building is maintained as an effective confinement boundary.
4.2 Ventilation Systems 4.2.1 Applicability Applies to ventilation systems established to support restricted activities.
4.2.2 Objective To specify surveillance requirements that will provide assurance that a ventilation system is operable when required.
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M i TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR i
4.2.3 Specifications 4
(1) At least once per week, whenever a ventilation system is required to be operating, verify :
that the direction of air flow is from zones oflesser potential for airbome contamination j
} to zones of greater potential for airbome contamination. '
1 (2) When a ventilation system is required to be operable, the exhaust air downstream of the j' filters shall be continecusly monitored or sampled to show that the specified 4
concentrations in 10CFR20, Appendix B are not exceeded.
1 ,
(3) Prior to placing a ventilation system in service, verify that all materials of construction
- for the verttilation systems are fire-resistant. All filters shall be verified to be of a fire-resistant type and, where applicable, listed by Underwriters' Laboratories or the l Fnctory Mutual Research Corporation. .
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(4) Prior to placing a ventilation system filter housing in service, verify that it includes an instrumentation device or multiple devices to indicate filter resistance and airflow rate. '
s i 4.3 Radiation and Effluent Monitorina Systems i
4.3.1 Applicability r l
This specification applies to the equipment and systems installed to detect radiation and/or contamination, e.g., laboratory counting instruments, and portable radiation measuring ihstrumentation used for the reactor facility.
4.3.2 Objective To describe check and calibration frequencies for laboratory counting instruments, and portable radiation measuring instrumentation. '
4.3.3 Specification I
(1) Upon initial acquisition, after major maintenance, and at least annually, stationary and portable monitoring instruments shall be calibrated using NIST traceable services.
(2) At least quarterly, background and efficiency shall be measured on all laboratory instruments used for counting health physics samples, using standard sources.
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TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR (3) Prior to placing an effluent monitoring instrument into service, after major maintenance, and at least annually thereafter while in service, a channel calibration of the ventilation effluent monitoring sampler and/or monitor shall be performed. At least monthly while in service, a channel check of the sampler or monitor shall be performed. These tests need not be performed if operation of the ventilation system is not required during the year.
4.3A Bases These specifications provide assurance that monitoring and analytical instrumentation will be functional when needed.
4.4 Effluents and Environmental Monitorinc
< 4.4.1 Applicability This specification applies at all times.
4.4.2 Objective To specify surveillance requirements to verify compliance with 10 CFR 20 requirements and to specify environmental monitoring requirements.
4.4.3 Specification 4.4.3.1 Air Effluent Surveillance Requirements Air effluent particulate monitors shall be examined at least once per week during physical handling of reactor vessel internal contents, the reactor vessel, biological shield, and during restricted activities to verify compliance with 10 CFR 20 limits. ;
l 4.4.3.2 Liquid Effluent Surveillance Requirements l
l Liquid effluents shall be sampled and analyzed prior to release, to determine whether they can be discharged directly or whether they require processing prior to discharge in accordance with the provisions of License No. SNM-770.
4.4.3.4 Environmental Monitoring Requirements The environmental monitoring requirements of License No. SNM-770 shall continue to be implemented during decommissioning activities at the WTR facility.
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TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- HTR 4.4.4 Bases The current programs for effluent and environmental monitoring for License No. SNM-770 include stationary and general air monitoring, weekly water monitoring at the weir discharge to Calley's Run, direct radiation monitoring, quarterly surface water runoff and stream measurements, quarterly drinking water sampling, annual sediment and soil sampling, vegetation sampling, and groundwater monitoring. These programs are comprehensive and appropriate for WTR decommissioning activities. A weekly analysis of air effluents is added to monitor for compliance with 10 CFR 20 requirements during restricted activities and during other decommissioning activities with the potential for creating airborne contamination that could be released io ilm environment.
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TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- HTR i l
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I.. -- 5.0 Desian Features '
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- j. See " Site Characterization" reports already referenced. The facility is located on the Westinghouse !i j
Waltz Mill Site which is owned and controlled by the Westinghouse Electric Corporation. The ;.
q approximate distance from the reactor building to the posted site boundary is about 200 yards. The
- restricted area as defined in 10 CFR 20 of the Commission's regulations shall be the containment j
- building. The controlled area as defined in 10 CFR 20 shall be the Central Operations Area of the Waltz Mill Site. '
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TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR i
6.0 Administrative Controls 6.1 Orcanization t i
6.1.1 The organization for the management and decommissioning of the WTR facility shall include l the following structure. Other organizational levels / staffing may be added to meet specific i facility needs.
(1) Level 1 - Individual responsible for the reactor facility licenses (i.e., General :
Manager, NSD).
(2) Level 2 - Individual responsible for the reactor facility activities (i.e., Waltz Mill Site !
Manager).
(3) Level 3 - Individual responsible for the day-to-day supervision (i.e., Radiation Safety Officer, other facility supervisors).
6.1.2 Responsibility Responsibility for the reactor facility shall be with the chain of command as specified in 6.1.1 above.
Individuals at the various management levels, in addition to having responsibility for the policies -
and activities conducted by the WTR facility, shall be responsible for safeguarding the public and ,
facility personnel from undue radiation exposures and for adhering to all requirements of the facility i license and technical specifications.
In all instances, responsibilities of one level may be assumed by designated alternates or by higher levels, conditional upon appropriate qualifications. I 6.2 Radiation Safety Committee 6.2.1 The Radiation Safety Committee shall provide management oversight and review of WTR decommissioning activities. The Radiation Safety Committee services to advise the Level 2 manager on matters that affect radiation safety, on areas where additional oversight or auditing is needed, and on items that involve an unreviewed safety question.
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TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR i
6.2.2 Charter and Rules i
Radiation Safety Committee activities shall be performed under a written charter or directive containing the following information, as a minimum. l t
A. Membership designation, including quorum requirements and provisions for alternates.
{
i B. Meeting frequencies.
C. Subjects reviewed, f i
D. Responsibilities.
E. Authorities.
(
F. Records.
G. Other matters as may be appropriate.
l 6.2.3 Review Requirements !
i The Radiation Safety Committee shall be responsible for review of the following:
A. Proposed activities that could affect personnel or facility safety or result in an uncontrolled release of radioactivity in excess of 10CFR20 limits, to be conducted without NRC approval, and reviewed and approved pursuant to 10CFR50.59 to verify l l the proposed activity does not constitute a change in the technical specifications or an l i
unreviewed safety question.
B. Proposed changes to the facility or to procedures required by Specification 6.4, that could affect radiation safety and that are to be completed without prior NRC approval l reviewed and approved pursuant to 10CFR50.59 to verify the activity does not constitute a change in the Technical Specifications or any unreviewed safety question.
C. All new procedures and revisions thereto the have significant effect on radiation safety.
D. Proposed changes to the Technical Specifications or the facility license.
4 A-16 l l l l I j
TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- H'TR E.
, Violations of the federal regulations, technical specifications, or facility license requirements.
F.
Unusual or abnormal occurrences which are reportable to the NRC under provisions of the federal regulations.
6.2.4 Audit Requirements Independent audits of decommissioning activities shall be performed under the cognizance of the Radiation Safety Committee. Audits shall include selective, but comprehensive, examination of activities, records, and documents with cognizant personnel, and observation of operations as appropriate. Audit personnel shall be technically qualified and should not have been involved in performance of the activity being audited. Audits shall include the following:
(1) Facility activities for conformance to the Technical Specifications and license, at least once per calendar year (interval between examinations not to exceed 15 months).
(2) The qualifications of the staff, at least once every other calendar year (inten al between examinations not to exceed 30 months).
(3) The results of action taken to correct those deficiencies that may occur in the reactor facility equipment, systems, structures, or methods of operations that affect facility safety, at least once per calendar year (interval between examination not to exceed 15 months).
Deficiencies uncovered that affect facility radiation safety shall immediately be reported to Level 2 management. A written report of the findings of each audit shall be submitted to Level 2 management and the manager of the radiation safety function within three months after the audit has been completed.
6.3 e Procedu_r_es Written procedures, including ALARA, shall be prepared and approved prior to initiating any activities listed in this section. Procedures for the following activities may be included in a single manual or set of procedures or divided among various manuals or procedures:
(1) Routine maintenance of major components or systems that could have an effect on facility radiation safety.
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TECIINICAL AND ENVIRONMENTAL SPECIFICA TIONS- IVTR (2) Surveillance tests and calibrations required by the Technical Specifications or those that may have an effect on facility radiation safety.
(3) Personnel radiation protection, consistent with applicable regulations.
(4) Administrative controls for maintenance and for the conduct of activities that could affect facility radiation safety.
Decommissioning activities shall be conducted utilizing the published and approved procedures except as noted below:
Substantive changes to the above procedures shall be made effective only after approval by appropriate management; changes which could affect radiation safety shall be reviewed by the Radiation Safety Committee. Minor modifications to the origiaal procedures which do not change their original intent may be made as a temporary change by Level 3 or higher and shall be documented; any temporary change that affects radiation safety must be reviewed by the Radiation Safety Committee within the following 45 days. All changes (except one-time deviations) shall be incorporated into the written procedures.
6.4 Experiments Review and Acoroval Experiments may be conducted without prior NRC approval if they have been determined to not involve an unreviewed safety question in accordance with the guidance in 10 CFR 50.59.
Experiments that affect radiation safety must also be reviewed by the Radiation Safety Committee.
6.5 Reauired Actions The following actions shall be taken in the event of an occurrence of the type identified in 6.6.2 (1)
- a. or 6.6.2 (1) b:
(1) Reactor facility conditions shall be returned to normal or the activities in progress stopped. Ifit is necessary to stop the activities in progress to correct the occurrence, operations shall not resume unless authorized by Level 2 or designated alternates.
5 (2) Occurrence shall be reported to Level 2 or designated alternates and to the NRC as required.
(3) Occurrence shall be reviewed by the Radiation Safety Committee.
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4 TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR r l 1- l 1
l . All reports shall be addressed to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555; l
Attention: Document Contro! Desk with a copy to the Regional Administrator Regio
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TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS - WTR 6.6.1 Annual Report Annually submit to the NRC a report containing the following:
(1) A nanative summary of facility activities.
(2) Tabulation of the major preventative and conective maintenance operations having safety significance.
(3) A brief description ofmajor changes in the reactor facility and procedures and activities significantly different from those performed previously and not described in the safety analysis report, and a summary of the safety evaluation that shows no l unreviewed safety questions were involved. :
l (4) A summary of the nature and amount ofradioactive effluents released or discharged i to the environs beyond the effective control of the licensee as determined at or before
)
the point of such releases or discharge. The summary shall include to the extent practical, an estimate of the majorindividual radionuclides present in the effluent. If I the estimated average release after dilution or diffusion is less than 25% of the ,
concentration allowed or recommended, a statement to this effect is sufficient. >
(5) A summarized result of the environmental survey performed outside the facility. J 6.6.2 Special Reports Special reports used to repon unplanned events as well as planned major facility or administrative l
changes shall be submitted in accordance with the following schedule. 1 (1) There shall be a report no later than the following working day by telephone and confirmed in writing by telegraph or similar conveyance to the NRC to be followed by a written report that describes the circumstances of the event within 14 days of any of the following:
- a. Release of radioactivity from the site above allowed limits (see 6.5):
- b. Any of the following (see 6.5):
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i TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- HTR (i) Activities in violation oflimiting conditions for conduct of activities established in the technical specification unless prompt remedial action is taken.
(ii) An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to facility operations.
(2) A written report within 30 days to the NRC of: i l
- a. Permanent changes in the facility organization involving Level 1 or 2 l personnel.
- b. Significant changes in the accident analysis as described in the decommissioning plan safety analysis.
6.7 Records Records may be in the form oflogs, data sheets, or other suitable forms. The required information may be contained in single or multiple records or a combination thereof. '
6.7.1 Records to be retained for a period of at least five years or for the life of the component involved ifless than five years:
(1) Normal facility operation (but not including supporting documents such as check lists, log sheets, etc., which shall be maintained for a period of at least one year).
(2) Principal maintenance activities.
(3) Reportable occurrences.
(4) Surveillance activities required by the technical specifications.
(5) Reactor facility radiation and contamination surveys where required by applicable regulations.
(6) Approved changes in operating procedures.
(7) Records of meeting and independent examination reports of the review and independent examination group.
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TECHNICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR i
6.7.2 Records to be retained for the lifetime of the facility:
NOTE: Applicable annual reports, if they contain all of the required information, may be used as records in this section.
(1)- Air and liquid radioactive effluents released to the environs.
4 (2) Off-site environmental monitoring surveys required by the technical specifications.
(3) Radiation exposure for all personnel monitored.
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(4) Drawings of the reactor facility. ;
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-(5) Records of disposal oflicensed material.
6.8 High Radiation Area 6.8.1 Parsuant to 10 CFR 20, in lieu of the " control device" or " alarm signal", each high radiation area, as defined in 10 CFR Part 20, shall be barricaded and conspicuously posted as a high _
radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation I Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Health Physics personnel) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 mR/h, prosided they are otherwise following plant radiation protection procedures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following. 1 l
- a. A radiation monitoring device which continuously indicates the radiation dose rate in I the area, or
- b. A radiation monitoring device which continuously integrates the radiation dose rate
. in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them, or
- c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for A-22
TECIINICAL AND ENVIRONMENTAL SPECIFICA TIONS- WTR providing positive comrol over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physics staffin the RWP.
6.8.2 In addition to the requirements of 6.8.1, areas accessible to personnel with radiation levels greater than 1000 mR/h at 45 cm (18 in) from the radiation source or from any -
surface which the radiation penetrates shall be provided with locked enclosures to prevent unauthorized entry, and the keys shall be maintained under the administrative control of health physics supervision. Enclosures shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in the area. In lieu of the stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may be inade by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.
For individual areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located within large areas, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the indisidual areas, then that area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a waming device whenever the dose rate in the area exceeds or will shortly exceed 1000 mR/hr i
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