ML20035B508

Winsor Commercial Nuclear Manufacturing Special Nuclear Matl License SNM-1067 Decommissioning Plan

ML20035B508
Person / Time
Site:	07001100
Issue date:	03/30/1993
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
To:
Shared Package
ML20035B506	List: ML20035B508 RA-93-003, Forwards Decommissioning Plan, Reflecting Scope of Activities Being Terminated Under SNM License SNM-1067
References
NUDOCS 9304020091
Download: ML20035B508 (83)

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Table of Contents 1.0 Introduction and General Information 1-1 1.1 Sde Description 1-2 1.2 Windsor Site Operations 1-2 2.0 Decommissioning Objectives, Activities, Tasks, and Schedules 2-1 2.1 Scope of Project 2-2 2.2 Decommissioning Objectives 2-2 2.3 Decommissioning Activities and Tasks 2-10 2A Decontamination Methodologies and Justifications 2-14 2.5 Procedures 2-16 I 2.6 Decommissioning Schedule 2-16 2.7 Decommissioning Organization and Responsibilities 2-18 2.8 Training 2-21 I 2.9 Contractor Assistance 2-21 3.0 Methods Used for Protection of Occupational and Public Health and Safety 3-1 3.1 Site Radiological History Building 17/21 Complex 3-2 3.2 ALARA 3-4 3.3 Health Physics Program 3-5 3.4 Contractor Personnel 3-14 3.5 Waste Management 3-14 4.0 Planned Final Status Survey 4-1 i Initial Issue i 3/30/93 -l

I Table of Contents 4.1 Final Status Survey Instrumentation 4-2 4.2 Methodology for Ensuring Adequate Survey Coverage 4-4 4.3 Methodology for Data Analysis and Comparison with Guideline Values 4-13 5.0 Physical Security Plan and Special Nuclear Material Control and Accounting Plan Provisions in Place During Decommissioning 5-1 5.1 Physical Security 5-2 5.2 Special Nuclear Material Control and Accounting 5-2 6.0 Funding 6-1 APPENDICES Appendix A: Uranium Characteristics A-1 I Appendix B: Temporary Storage of Radioactive Material B-1 Appendix C: Glossary C-1 l Appendix D: References D-1 l I I I I Initial Issue ii 3/30/93 I

List of Ficures 1-1) Combustion Engineering Location Within Connecticut 1-3 1-2) Combustion Engineering Location - Local Area 1-4 2-1) Windsor Commercial Nuclear Fuel Manufacturing Complex, General Area 2-3 2-2) Windsor Commercial Nuclear Fuel Manufacturing Complex 2-4 2-3) Building 17 Plan View 2-5 2-4) Pellet Shop Plan View 2-6 2-5) Pellet Shop Mezzanine Plan View 2-6 2-6) Pellet Shop Section View - West Detail 2-7 2-7) Pellet Shop Section View - Center Detail 2-8 2-8) Pellet Shop Section View - East Detail 2-9 2-9) Windsor Nuclear Fuel Manufacturing Decontamination and Decommissioning Schedule 2-17 2-10) Windsor Nuclear Fuel Manufacturing Decontamination and I Decommissioning Organization 2-20 3-1) Sample Radiological Survey Form 3-13 4-1) Standard Measurement / Sampling Pattern For Systematic Grid Survey of Structure Surfaces 4-6 4-2) Grid Demarcation of Pellet Shop Ceiling 4-7 4-3) Soil Sampling Locations 4-11 A-1) Uranium Decay Series for Radionuclides of Concern A-4 I initial issue iii 3/30/93 I P

I L.ist of Tables 4-1) Decommissioning Release Criteria 4-3 4-2) Factors for Comparison of Survey Data with Guidelines and Determining Additional Data Needs 4-17 .I A-1) Principle Uranium and Uranium Progeny Radiations A-5 I I I i l 1 I I l I I I i Initial Issue iv 3/30/93 I

I I I. I I I 4 I SECTION 1 I I I I INTRODUCTION AND GENERAL INFORMATION j g I o I O l I u I i Initial Issue 1-1 3/30/93 l I _--_-____.____.---.__._________--)

I 1.0. Introduction and General Information: This Decommissioning Plan provides the general framework for decontaminating and decommissioning the Combustion Engineering, Inc. (C-E) Windsor commercial nuclear fuel manufacturing complex I (figure 2-2). The plan describes the scope of the project, including but not limited to, site history, extent of related contamination, objectives of remediation activities, the methodologies that will be used to achieve those objectives, survey techniques that will be used to verify that the release criteria have been met, and documentation that will be maintained to establish a completion record. The plan begins with the general history of the fuel manufacturing site, lists the objectives of the project, and details the I decontamination and decommissioning methodologies and activities planned for the commercial manufacturing complex. The plan outlines the p'anned final status survey and shows details of administrative, security, training, accountability, and health and I safety activities associated with the project. Until this Decommissioning Plan is approved, Combustion Engineering will conduct decontamination activities in accordance with the applicable requirements contained in Special Nuclear Materials license SNM-1067. 1.1. Site

Description:

The Combustion Engineering Windsor site is located in the north central portion of Connecticut (Figure 1-1). The site is on approximately six hundred acres along the section of the Farmington river known as the Rainbow Reservoir in the town of Windsor, Connecticut (Figure 1-2). 1.2. Windsor Site Operations: The commercial nuclear fuel manufacturing history of the Windsor site dates from late 1968, when SNM-1067 was issued to Combustion Engineering. Commercial activities covered by SNM-1067 have involved development and production of fuel products for the commercial nuclear industry. Initially, uranium dioxide fuel pellets purchased from an outside supplier were processed and loaded into fuel rods at building 17, the manufacturing facility at the Windsor site. In 1970, a fuel pellet operation was added to the already existing fuel pellet loading and fuel assembly operation. At that time, building 17 began to receive uranium dioxide (UO ) powder, 2 which was subsequently pressed into fuel pellets. UO pellet pressing and powder 2 handling operations continued until permanently halted in December of 1989. Upon ti:e I cessation of powder operations, a major decontamination project was undertaken in the pellet shop of building 17, resulting in greatly reduced contamination levels in the manufacturing facility. Since that time, manufacturing activities in building 17 have I been limited to producing fmished fuel assemblies from sintered UO, pellets manufactured at and received from another site. Current uranium bearing fuel manufacturing operations are scheduled to cease on or about September 30,1993. Immediately following the cessation of uranium handling I operations, it is the intention of Combustion Engineering to proceed with the decontamination and decommissioning of the building 17/21 areas covered by SNM-1067 and used for uranium bearing fuel production and storage activities, as described in this plan. Initial Issue 12 3/30/93 i I l __-__-_ - ____ --___-__-_-_-__--_-_-__ _ ______ _ ____-________-____-_ - ______.___________ D

l Figure 1-1 l Combustion Engineering Location within Connecticut 7 l Combustion Engineering,Inc. Windsor, Connecticut i t MASSACHUSETTS 5 y.-- w l t HARTFORD 1 RHODE NEW g ISIAND l YORK w y / l t 5 i l / sanxirront LONGISIAND SOUND i e STAWORD I l A J .l North t (Not to Scale) l Initial Issue 13 3/30/93 i i

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I I l I I i SECTION 2 g E l-l DECOMMISSIONING OBJECTIVES, g ACTIVITIES, TASKS, AND SCHEDULES I 4 I I LI I i Initial Issue 2-1 3/30/93 I

I 2.0. Decommissioning Objectis es, Activities, Tasks, and Schedules 2.1. Scope of Project: The decontamination and decommissioning project will encompass I the following facilities (see Figures 2-1 through 2-8 for locations oflisted areas): 2.1.1. Building 17: This building is the location of commercial fuel manufacturing I activities under SNhi-1067. 2.1.2. Building 21: This warehouse was used for storage of packaged radioactive I materials. 2.1.3. Land Areas Surrounding Buildings 17 and 21: These areas were the site of [ I radioactive material storage in support of shipping and receiving activities. 1 Although radioactive material is not stored within 10 feet of the fence, for the I purposes of surveying, these land areas are considered to extend to one meter beyond the fence line. 2.1.4. Drainage Swales to the South and West of the Building 17/21 Complex: These areas were the recipients of rainwater runoff from the building 17 roof top, paved apron, and parking lot. 2.2. Decommissioning Objectives: The goals of the Combustion Engineering Windsor Nuclear Fuel hianufacturing Decontamination and Decommissioning project are as follows: 2.2.1. Perform tasks in a safe and environmentally acceptable manner in accordance l with applicable local. state, and federal regulations. 2.2.2. hiaintain exposures to radioactive material As Low As Reasonably Achievable (ALARA). 2.2.3. hiinimize the volume of radioactive waste generated. 2.2.4. Decontaminate the building 17/21 complex as required to meet the criteria specified in Table 4-1 and free release the building for unrestricted use. 2.2.5. Verify that soils inside and up to one meter outside the building 17/21 complex fence line, as well as the two drainage swales, meet the criteria shown in Table I 4-1 and are available to be free released for unrestricted use. 2.2.6. Verify by survey that applicable criteria for release have been met. Initial Issue 2-2 3/30/93 g

Figure 2-1 Windsor Commercial Nuclear Fuel Manufacturing Complex General Area I (Map not to scale) I N f S w ale,1 i / l l m x l '%. _ / o' g ,~~ l l a' N,~! l \\ 1 a' l e l ,/ suieing ! PARKING / 21- [ e,~,' a e 8'* N., \\ N I Building I N 17 , N. l . ~. ( s e / s \\ s N, / Swale 3 '., ,/ \\ H-l'/ e I '~--l ()H-2 1 ()H-3 OH-4 I COMBUSTION ENGINEERING Windsor Commercial Nuclear Fuel Manufacturing Complex General Arca 4 I l LEGEND 4 DENCTES HOTWASTEUNE North AND MANHOLEIDCATON / H-# DENOTES VANHOLE N'JMBER j

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q l i l \\ l 2.3. Decommissioning Activities and Tasks 2.3.1. Area Survey Assessment: The survey effort that will be applied to the various areas included in the project will vary according to their usage history. (See Figures 2-1 through 2-8 for locations oflisted areas.) The emphasis ofinitial surveys will be to better define the extent of the remediation effort required in each area. Surveys will be performed as follows: 2.3.1.1. Building 17 Pellet Shop: During fuel fabrication, the pellet shop was I surveyed on a routine basis and the quantity and extent of contamination in the shop is well known. Additionally, as part of the I radiation work permit program as described in section 3.3.2, prior to i beginning each task, a pre-job survey will be conducted. Following 4 completion of decontamination efforts, these areas will be surveyed in I accordance with the final status survey plan (see section 4) to demonstrate that they meet the criteria listed in Table 4-1. 2.3.1.2. Building 21: Building 21 is a warehouse used in support of fuel production activities. As no contamination is anticipated, the building will be surveyed in accordance with the final status survey plan (see section 4) to demonstrate that it meets the criteria listed in Table 4-1. This plan calls for the unrestricted release of building 21 following the cessation of uranium operations. Once the final status survey for building 21 is accepted by the NRC, the building may be turned over for other uses. 2.3.1.3. Land Areas inside and Within One Meter of the Building 17/21 Complex Fence Line: As no contamination is anticipated in these areas, they will be surveyed in accordance with the final status survey I plan (see section 4) to demonstrate that they meet the criteria listed in Table 4-1. 2.3.1.4. Drainage Swales to the West and South of the Building 17/21 Complex: Rainwater runoff from the roof of building 17, the apron - g around building 17, and the building 17/21 complex parking lot E drained through these areas. Scoping surveys performed and documented during the period 1988 through 1991 and sample results from those surveys indicate that further characterization and/or remediation are not warranted. These areas will be surveyed in accordance with the final status survey plan (see section 4) to demonstrar that they meet the criteria listed in Table 4-1. 2.3.2. Planned Decommissioning Activities: The following section details the decontamination activities planned for each area designated as within the scope Initial Issue 2-10 3/30/93 1

of this project. (See Figures 2-1 through 2-8 for locations of specific areas listed.) Area Activity Building 17: Pellet Shop This area was the subject of a. major I Revitalization Area (Column decontamination effort in late 1990 and early 1991 location 4-8, A-C) (Figures 2-3 (see section 3.1.2). This project coincided with through 2-5 and 2-7) the cessation of pellet pressing operations at the site. Since that time, contamination levels have remained minimal. The interior structure fabricated during the decontamination project will I be removed and decontaminated. Removal of the materials comprising this structure will allow the decontamination of underlying materials and will facilitate access to the building 17 structure for the final status survey. The area will be decontaminated as required and will be surveyed in accordance with the final status survey plan (see section 4) to demonstrate that it meets the criteria listed in Table 4-1. When survey results demonstrate that these criteria have been met, the area will be free released and made available for unrestricted use. Building 17: RP Office, Ilydrogen The Radiation Protection count room will be Room, Change Room Complex relocated to the NE corner of the building (east of (Column location 12-14, A-C) column 14, A-B) to facilitate the dismantling of (Figures 2-4 and 2-8) the pellet shop. The change room will be realigned in order to improve ingress to and egress from the areas being decontaminated. The women's change room and the hydrogen room structures will be dismantled and the materials processed for disposal in order to facilitate the i final status survey. Structural materials will be removed, decontaminated, and surveyed to l demonstrate that they meet the criteria listed in-l Table 4-1. When survey results demonstrate that I these criteria have been met, the materials will be free released and made available for unrestricted use. 1 I Initial Issue 2-11 3/30/93

I Area Activity Building 17: Stack and Load Room The stack and load room will be dismantled to I (Column location 8-11, A-B) allow for more efficient decontamination of 6e (Figures 2-3 and 2-4) materials comprising the room and to assure better access to building 17 structures during the final I status survey. Removed materials containing residual radioactivity will be decontaminated and surveyed to demonstrate that they meet the criteria I. listed in Table 4-1. When survey results demonstrate that these criteria have been met, the I materials will be free released and made available for unrestricted use. Building 17: Plumbing and Plumbing and electrical equipment will be Electrical Conduit in Pellet Shop removed as it becomes available. Removal of these items will allow improved access to building g l 17 structures for decontamination and for the final l status survey. Removed materials will be decontaminated and surveyed to demonstrate that they meet criteria listed in Table 4-1. When survey results demonstrate that these criteria have been met, the materials will be free released and l made available for unrestricted use. lg Building 17: Ventilation Ductwork Ventilation ductwork will be removed when it is [ in the Pellet Shop no longer required to maintain essential services. Removal of these items will allow better access to building 17 structures for decontamination and the I final status survey. Removed materials will be decontaminated and surveyed to demonstrate that they meet the criteria listed in Table 4-1. When survey results demonstrate that these criteria have been met, the materials will be free released and made available for unrestricted use. Building 17: Ilot Waste Piping 1101 waste piping under building 17 will be I. (Within Building 17 and the Line removed from the ground and decontaminated. Running from Building 17 to the The line running from building 17 to the junction Junction inside Manhole # 11-4) inside manhole number 11-4 will be surveyed at (Figure 2-1) normal access points. If survey results meet the values listed in Table 4-1, piping will be left in place. If survey results indicate contamination in excess of the values listed in Table 4-1, efforts l Initial Issue 2-12 3/30/93 'I

I Area Activity will be made to decontaminate piping by flushing. I If decontamination efforts are not s ccessful, piping will be removed. By project completion. the building 17 hot waste piping connection will I be severed. Building 17: Balance of the Pellet Structural materials will be decontaminated as Shop (Figures 2-2 through 2-8) required to meet the criteria listed in Table 4-1. Materials and equipment remaining after cessation I of manufacturing operations will be decontaminated to meet the criteria listed in Table 4-1 and may be removed from the facility. Floor I surfaces will be decontaminated to meet the criteria listed in Table 4-1. As much of the physical plant will be left intact as practicable. Decontamination methodology will be a combination of the techniques described in section 2.4 below. When survey results demonstrate that material, equipment, and structures meet the criteria listed in Table 4-1, they will be free released and made available for unrestricted use. Building 21: Warehouse The warehouse has been used to store packaged (Figures 2-1 and 2-2) radioactive materials, and is not expected to be contaminated. Building 21 will be suneyed in accordance with the final status survey plan (see section 4) to demonstrate that it meets the criteria listed in Table 4-1. When survey results demonstrate that these criteria have been met, the building will be free released and made available I for unrestricted use. Land Areas Inside and Within One These areas may have been subject to rainwater I Meter Beyond the Building 17/21 runoff and/or ventilation effluent fallout and have Complex Fence Line and Swale been sampled as part of the environmental I Areas West and South of the Fence monitoring program. Based on this program, it Line (Figures 2-1 and 2-2) does not appear that these areas are contaminated, As such, they will be surveyed in accordance with the final status survey plan (see section 4) to i demonstrate that they meet the criteria listed in Table 4-1. When survey restdts demonstrate that g these criteria have been met, the swales will be initial Issue 2 13 3/30/93 ,!I

Area Activity free released and made available for unrestricted use. Buildings 17 and 21 Exterior Walls These areas are not likely to be contaminated. and Building 21 Roof: They will be surveyed in accordance with the final status survey plan (see section 4) to demonstrate that they meet the criteria listed in Table 4-1, When survey results demonstrate that these criteria have been met, these areas will be free released and made available for unrestricted use. Building 17 Roof: Potential contamination on the roof will be I characterized. Following remediation. if such is required, this area will be surveyed in accordance with the final status survey plan (see section 4) to demonstrate that it meets the criteria listed in Table 4-1. When survey results demonstrate that these criteria have been met, the area will be free released and made available for tmrestricted use. 2.4. Decontamination Methodologies and Justincations: It is the intention of Combustion Engineering to complete required decontamination activities using a combination of technologics. Following is a summary of the primary decontamination technologies that will be used as appropriate for decontamination activities in areas I within the scope of this project: 2.4.1. Hand Wiping: It is anticipated that the majority of the contaminated equipment I and materials will be cleaned using hand wiping. This will involve cleaning equipment, structure, and/or material surfaces manually using rags and/or abrasive pad: and a cleaning solution such as Alconox or similar. I Decontamination solutions will be carefully selected to ensure that mixed waste is not generated. This method has been selected because it is a simple and effective method for routine decontamination tasks and personnel are easily l I trained to safely use this technique. Rags used in these operations may be laundered and reused.

2.4.2. 11ydrolasing

This methodology is an available option for removal of widespread areas ofloose surface contamination. This method will only be I utilized in an enclosed area that is set up specifically for hydrolasing. Specific j allowances will be made for the recapture and recycling of contaminated water. Initial Issue 2 14 3/30/93

t I 2.4.3. CO Cleaning: This technology is an available option for surface cleaning by 2 means of tiny, solidified pellets of CO. The pellets sublimate when contacting 2 the surface of the contaminated material, in effect sweeping the contaminants I off the surface in a rush of gaseous CO. The gaseous CO is then collected by 2 2 a suction line and passes through a IIEPA Blter. Advantages to this process include its effectiveness at cleaning intricate surfaces, its lack of physical I, surface damage, and its lack of secondary radioxtive waste generation. 2.4.4. Strippable Paint: This material, commonly used in nuclear power plants, may have some applicability in two areas. First, it may be used to provide a removable, protective coating over decontaminated areas remaining in the I vicinity of ongoing decontamination work. Second, it may be utilized to perform actual decontamination tasks on large smooth surfaces such as metal sheets or similar. If used, it is intended that, following its removal, the paint I-may be incinerated in an approved facility, thus reducing radioactive waste burial volumes. I 2.4.5. Soil Removal: Soils with contamination exceeding the values in Table 4-1 and section 4.2.22.1 of this plan will be removed and transported to an approved j disposal facility. Soils that are excavated may be further segregated and/or categorized to separate soils that are free releasable from soils requiring disposal. Additionally, soils may be cleaned of contaminants where practical. 2.4.6. Abrasive Media Cleaning: This process utilizes either sponge-like material, sand, or grit accelerated by high pressure air to effect controlled decontamination of contaminated surfaces. These media have the advantage of being reusable. When exhausted, the sponge material may be incinerated in an approved facility, resulting in relatively small volumes oflow level waste. The sand and/or grit will be disposed of as low level radioactive waste if it exceeds i the criteria listed in Table 4-1. 2.4.7. Material Surface Removal: In cases where hand wiping or other surface I-cleaning methods are insuf6cient to remove contaminants present, removal of material surfaces may be required. For tools, equipment, and other metal objects, grinders may be utilized to remove contaminants that would otherwise be considered Exed contamination. This will avoid the necessity of disposing of large inventories of materials as radioactive waste. For contaminated structural materials such as concrete floors and walls, scabblers and scarifiers may be utilized to remove contaminated surfaces from those materials. To the extent practical, engineering controls will be used to reduce aerosols that can result from scabbling and/or scarifying operations. Scabbling has been proven successful in a number of areas and buildings during recent decontamination projects. Initial Issue 2-15 3/30/93

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2.5. Procedures

Combustion Engineering will conduct decontamination and decommissioning work using written procedures approved by the Program Manager, Decontamination and Decommissioning. _ Decommissioning tasks in radiologically I controlled areas may also be controlled by radiation work permits (section 3.3.2). Additionally, for major decontamination and decommissioning tasks, the procedures and/or radiation work permits may be supplemented by task plans. 2.5.1. Task plans include information about both precautionary and performance related issues. Task plans emphasize ALARA as well as compliance with applicable radiological and industrial safety requirements and will help assure that radiological controls remain a high priority. 2.6. Decommissioning Schedule: The planned schedule for the decontamination and decommissioning portion of the Windsor decommissioning project is shown in Figure I 2-9. Following completion of decontamination and decommissioning activities by the fourth quarter of 1995, work is expected to begin on final status surveys. One exception is the final status survey on building 21 and immediately surrounding areas (as discussed in section 2.6.1 below). Final status surveys are planned according to the following schedule: l 2.6.1. Building 21 and immediately surrounding areas: The survey will commence on or about September 30,1993, with the intent of making building 21 and the areas immediately surrounding building 21 available for unrestricted use by the l second quaner of 1994. 2.6.2. Building 17 pellet shop, roof, exterior surfaces, and clean areas; building 17/21 I complex exterior land and paved areas; and the hot waste piping from building 17 to the junction in manhole number H-4: These surveys are scheduled to commence on or about October,1995, following completion of { decontamination and decommissioning tasks and activities. 2.6.3. Final Status Survey Report: Preparation of the Final Status Survey Report is ) I expected to commence following completion of final status surveys. Completion of the repon is scheduled to take approximate six months. I I I I initial Issue 2-16 3/30/93 I

uma e sus um a suus e m m m m m m m Figure 2-9 g Windsor Nuclear Fuel Manufacturing Decontamination and Decommissioning Schedule =- E' 1993 1994 1995 g r> - > o c o ' ' >c -a a- >o = n ' ' >c - o a-e * " * ' s = o o Activi'Ly-Task 02 0 7 a " * * ' s e o

• 5

< m0Z 0 ,a a 2 < 2 ,a a 2 < 2 ,o < m0 Decontamination Set-up Decon Area i and - - - - - - - - - - - - - - - - - - - ~ ~ - - - DOCommiSSIOnin9 Remove Equipment Activities Remove HEPA Ventilation Units and Ductwmk Remove HVAC Units and Ductwork a Decontaminate Cei!ing m m Remove Interior Structures Remove interior Fixtures Deconta ninate Interior Surfaces Decontaminate Floors u, Q Remove Hot Waste Unes . ~ _ _ _,.. _.

i l 2.7. Decommissioning Organization and Responsibilities: This section describes the organization of the Windsor decommissioning project. Positions within the organization with responsibilities related to decommissioning safety are identified and their functions I described. The minimum qualifications for these positions are also described. Figure i 2-10 depicts the organization structure. The positions described in sections 2.7.1 l through 2.7 3 will be filled by full time personnel. Services described in sections 2.7.4 I through 2.7.7 are functional in nature and will be provided on an as needed basis. 2.7.1. Program Manager, Decontamination and Decommissioning: The Program I Manager, Decontamination and Decommissioning reports to the Vice President, Regulatory Affairs. The Program Manager has the overall responsibility and I authority for the safe conduct of decontamination and decommissioning activities. lie or she is responsible for approving programs, procedures, and plans. 2.7.1.1. The minimum qualifications for this position are a bachelor's degree in one of the sciences or engineering with four (4) years experience, including at least one (1) year in a management position in the nuclear industry. 2.7.2. Project Manager, Decontamination and Decommissioning: The Project Manager, Decontanination and Decommissioning reports to the Program Manager, Decontamination and Decommissioning. Ile or she is responsible for overseeing the planning and execution of day to day decontamination actisities for the project. The Project Manager creates, monitors, and modifies work plans and schedules as required. 2.7.2.1. The minimum qualifications for this position are a bachelor's degree in one of the sciences or engineering and three (3) years experience 1 I in health physics, decontamination, or other related field. Alternately, l a minimum of ten (10) years experience in health physics and/or decontamination, with two (2) years in a supenisory capacity. I l 2.7.2.2. The Project Manager, Decontamination and Decommissioning is responsible for supenising the decontamination team. The decontamination team performs the majority of the decontamination tasks associated with decommissioning building 17 including cleaning I structures and materials, and dismantling structures and equipment as required. I 2.7.3. Radiation Safety Officer: The Radiation Safety Officer reports to the Program Manager, Decontamination and Decommissioning. The Radiation Safety Officer is responsible for defining and implementing procedures related to radiological safety. The procedures address safety criteria, monitoring, and initial Issue 2-18 3/30/93 I

I i training necessary to ensure the protection of employees, the public, and the environment. As part of this responsibility, he ensures that ALARA is considered in the decontamination process. I i The Radiation Safety Officer has no production responsibility. If the Radiation Safety Officer believes an operation to be unsafe, he or she has the authority to I halt that operation. Operations halted for safety reasons shall not be restarted without the concurrence of the Radiation Safety Officer or the Program Manager, Decontamination and Decommissioning. 2.7.3.1. The minimum qualifications for this position are a bachelor's degree I in one of the sciences or engineering, with two (2) years experience in health physics. i 2.7.3.2. The Radiation Safety Officer is responsible for supervising the radiological protection team. The radiological protection team performs radiological surveys, air sampling, and radiological and industrial safety job coverage for tasks associated with the decommissioning project. The Decontamination and Decommissioning Organization will also draw support in the following disciplines from other members of the Combustion Engineering staff: 2.7.4. Radioactive Waste: The processing, packrging, and shipping of radioactive waste from the deccmmissioning operations. 2.7.5. Industrial Safety: Assistance will be provided relative to the industrial safety aspects of the project. i

2.7.6. Licensing

Ser ices will be provided to interface with the U.S. NRC and coordinate licensing issues related to decontamination and decommissioning.

2.7.7. Radio-chemistry

Services will be provided for performance of sample analyses in support of Decontamination and Decommissioning activities. In addition to the above, personnel from the large pool cf highly experienced engineers and scientists within the Combustion Engineering organization will be used to' provide specialized support as required. I I i Initial Issue 2-19 3/30/93 I I

uma uma men num seu muu ums susu amm num num sums uma sums ums-uma uma muu man Figure 2-10 y i WINDSOR NUCLEAR FUEL MANUFACTURING [ DECONTAMINATION & DECOMMISSIONING ORGANIZATION Vice President, Regulatory Affairs Program Manager, y Decontamination & Decommissioning O Radiation Safety Project Manager Licensing Officer Decontamination industrial Safety Radiochemistry as Prot tn b

2.8. Training

Training for the Windsor decommissioning project will be provided commensurate with the hazards faced by the worker. The training program will define training requirements for in-house workers, contractors, and visitors requiring unescorted facility access. 2.8.1. The training is provided to enable personnel working in the facility to perform I their tasks in a safe manner, without endangering themselves, other Combustion Engineering personnel, the facility, the public, or the er vironment. 2.8.2. Appropriate training will be provided in accordance with a training program to personnel who work within or visi< the facility. The degree of training provided each individual will be commensurate with the extent of the radiological, chemical, and/or industrial hazards likely to be encountered. 2.8.2.1. Topics that will be included in personnel training programs include but are not limited to. procedures for the storage, transfer, and use of I radioactive materials; health protection problems associated with exposure to the types of radioactive materials and radiation that will be encountered during the project; precautions or procedures to minimize exposure (ALARA); the purposes and functions of protective devices and/or equipment employed in the building 17/21 complex; the need to observe, to the extent within the worker's control, the applicable provisions of Nuclear Regulatory Conunission regulations and licenses for the protection of personnel from exposures to radiation and radioactive material; the responsibility of the worker to promptly report to the Combustion Engineering management any condition which may lead to or cause a violation of Nuclear Regulatory Commission regulations and licenses or unnecessary exposure to radiation or to radioactive material; the appropriate response to warnings made in the event of an unusual occurrence or malfunction that may involve exposure to radiation or radioactive material; and the radiation exposure reports that which I workers may request (as referenced in 10 CFR f 19.13, as amended through 5-31-91). 2.8.2.2. Escorted visitors will be exempt from the training requirements. p 2.8.3. Formal training will be documented and the documents retained for two years .6 or the duration of the individual's employment, whichever is greater. 2.9. Contractor Assistance: Combustion Engineering may choose to accomplish some decontamination and decommissioning activities by using contractors; however in the case of activities performed by contractors, the responsibility for safety and compliance remains with Combustion Engineering. It is not anticipated that decontamination and Initial issue 2-21 3/30/93 I -____-__-_____-____-____-__--_________-___-_________-___--_-_a

I decommissioning activities will be performed under subcontractor licenses. Areas that may require contractor assistance include, but are not limited to, heavy equipment and construction support for tasks involving structural alterations and/or excavations. Contractors providing senices for the decontamination and decommissioning project will be subject to the same training and procedural compliance requirements as Combustion Engineering personnel. There will be no distinction in the applicability of radiological and/or industrial safety controls between contractors and permanent staff. l l t I i I I i I 1 r I I i I i 6 I I Initia! Issue 2 22 3/30/93 I

I 1 I I I i I SECTION 3 i I i I i I METHODS USED FOR. PROTECTION OF I OCCUPATIONAL AND PUBLIC HEALTH l l AND SAFETY I i I -l t i Initial lssue 31 3/30/93 i - I 1

L 3 0. Methods Used for Protection of Occupational and Public IIcalth and Safety: This section contains a description of the methods that will be used to ensure protection of l5 w rkers and the environment against radiation hazards during decommissioning. It 1 ll addresses both administrative and technical issues related to safety, as well as the operational history of the building 17/21 manufacturing complex. 3.1. Site Radiolcgical History,Iluilding 17/21 Complex: Commencement of uranium handling operations in building 17 coincided with the issuance of SNM-1067 late in 1968. The facility was divided into a " clean" area (with no radioactive contamination 'I hazards) for manufacture of fuel assembly hardware and a " hot" area (with complete radioactive contamination controls in place) for uranium fabrication work. Building 21 l was operated as a clean facility, sersing as a materials warehouse for production l supplies and packaged radioactive materials. The land areas inside the fence were i maintained clean and utilized in support of shipping and receiving activities. The material involved in fuel production was then, and remains today, limited to uranium I enriched in the *U isotope to a maximum of five percent by weight. Comm"nial operations have involved receipt of UO in pellet or powder form, pressing of the 2 powder into pellets, and loading of the pellets into fuel rods that were assembled into fuel bundles. General records indicate that minor spills occurred occasionally in the pellet shop, were generally associated with production operations, and were promptly cleaned up. There are no records of spills or accidents with the potential to impact the health and safety aspects of the decommissioning project. !u December,1989, operations with UO powder ceased permanently and an extensive decontamination 2 project was undertaken. (Sections 3.1.1 and 3.1.2 below detail the two phases, l redeployment and revitalization, of the project.) (Figures 2-3, 2-4, 2-5, and 2-7 show the location of the referenced areas.) As e result of this project, UO contamination 2 levels in the shop were dramatically reduced. In its current configuration, the pellet shop is only minimally contaminated with UO - 2 l

3.1.1. Redeployment

The December,1989 cessation of uranium powder pressing and handling operations in building 17 necessitated the large scale relocation of i ' g pellet pressing and associated support equipment no longer needed in Windsor. 3 This was begun in December,1989 and continued through the later part of l 1999. During this process, equipment removed included pellet presses, lJ sintering furnaces, powder preparation stations, a micronizer, powder l E preparation belts, an oxidation and reduction furnace, dewaxing furnaces, j blenders and blender hoods, and miscellaneous other equipment used in support of operations for processing UO powder into pellets. A portion of the press 2 mezzanine was also removed from the pellet shop. The ventilation system t l. underwent considerable modification as well. Approximately 20% of the l: - HVAC ductwork in the pellet shop was replaced as well as approximately 50% of the HEPA ventilation system ductwork. Additionally, ventilation systems FA-2 and FA-4 were also removed. As the equipment was removed, there was a significant decontamination effort within the pellet shop. initial Issue 32 3/30/93 1 1 I

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3.1.2. Revitalization

A project was begun with the aim of readying the shop for the planned installation of a new, semi-automated fuel rod loading line. (That installation plan was subsequently cancelled.) A primary goal of the project I was decontamination of the areas formerly occupied by the pellet pressing operation to allow for the clean installation of the new semi-automated rod line. To that end, a decontamination effort commenced in late 1990 and lasted through July,1991. During that time, the shop was decontaminated from the double doors at the west end of the pel:et receiving area west to the annex roll-up door (column lines one to eight). (See Figures 2-3 through 2-7 for I depictions of revitalization area and pellet shop annex.) The area affected by the project amounted to approximately sixty percent of the total square footage of the pellet shop. The decontamination effort encompassed walls, floors, ceilings, and fixtures. Initial removable alpha contamination levels ranged from approximately 2,000 2 to 10,000 dpm/100 cm in the general work area, with levels greater than 2 10,000 dpm/100cm in specific isolated areas such as equipment interiors and other surfaces not routinely accessed. Scaffold was erected to provide access to the upper shop areas. The decontamination team consisted of contractor decontamination technicians operating under the direction oflicensee l management. The decontamination effbrt, which lasted approximately six weeks, included hand wiping the walls, ceilings, and associated support structures, and scabbling the concrete floors. Work commenced initially in the ceiling areas in the interest of maintaining a logical task progression that minimized the potential for recontamination of surfaces previously decontaminated. The job successfully progressed from the overhead to the walls and then to the floors. Scabbling of the floors proved quite successful at reducing contamination levels on the concrete. Two different scabbler types i were used; a large scale 5 head unit was used for open floor areas while g 3 detailed areas such as corners and wall to floor joints were decontaminated with hand held single head units. The scabblers were used to remove approximately g 1/8 inch to 1/4 inch of the contaminated floor surface. The floor 3 decontamination experience demonstrated that contamination has not significantly penetrated the pellet shop floor. Upon completion of the decontamination project, the remainder of the existing press mezzanine was relocated twenty feet to the east and a prefabricated structure covering approximately thirty percent of the pellet shop was installed between the mezzanine and the double doors of the pellet receiving area to act as a clean lI buffer room for part of the semi-automated rod line. Although the semi-automated rod line installation was cancelled, the l decontamination effort undertaken during revitalization proved quite successful. Final surveys taken at the end of the project show removable alpha i 2 2 contamination levels averaging between 100 dpm/100cm and 300 dpm/100cm, Initial Issue 3-3 3/30/93

g l 5-

2 (maximum of 1,000 dpm/100cm ) in the shop areas which had received only a single wipedown pass (outside the revitalization area). Inside the prefabricated structure, where more intensive decontamination efforts were undertaken, 2 removable alpha contamination levels werc <50 dpm/100cm. Monitoring of the decontamination team using personal lapel air samplers showed that the radiological controls practiced during the project were effective. The work crew averaged approximately 0.6 MPC hours per person per work day (prior to taking credit for respiratory protection), well below both regulatory and administrative limits. The removal of significant amounts of contamination I during revitalization greatly reduced the potential hazard workers will face during this project. It also provided the decontamination and decommissioning team direct experience with large scale decontamination projects. 3.1.3. Current Data, Pellet Shop: In the time since revitalization concluded, work in the pellet shop has been confined to operations involving only sintered UO2 pellets. The lack of powder operations has resulted in the contamination levels in the shop remaining minimal. The generally accessible areas of the shop l have removable alpha contamination levels averaging <500 dpm/100cm. (Less 2 accessible areas and the insides of equipment such as ventilation hoods and the stack and load table are higher but are restricted to small, isolated areas.) I 3.1.4. Current Data, Balance of Building 17/21 Complex: Building 21, the clean shop areas of building 17, and the land and paved areas inside the restricted area l fence have continued to be used as originally designated, functioning as clean areas used in support of component fabrication, materials storage, and shipping and receiving activities. Surveys are performed in those areas monthly. Review of the survey data generated since January,1992 indicates that contamination levels in these areas are insignificant.

3.2. ALARA

The decommissioning team has a strong coLaitment to the ALARA philosophy. As such, a key objective of the project is to maintain exposure to radioactive material ALARA for the public, the environment, and workers at the Windsor site. The following policies will be an ongoing part of the decommissioning process: 3.2.1. ALARA targets will be set and trends will be monitored and recorded during the decemmissioning project. The indicators that may be used include, but are not limited to: 3.2.1.1. Liquid Effluent: A measure of the amount of uranium released in liquid effluent. 3.2.1.2. Air Efiluent: A measure of the amount of uranium emitted from building 17 via the ventilation system, monitored at the exhaust. 1 Initial issue 3-4 3/30/93 ) I

I 3.2.1.3. Shallow Dose Equivalent: Shallow Dose Equivalent is the external dose to the skin at a tissue depth of 0.007 cm. 3.2.1.4. Total Effective Dose Equivalent (TEDE): TEDE is the sum of the deep dose equivalent and committed effective dose equivalent. 3.2.1.5. Airborne Radioactivity: A measure of the concentration of radioactivity in the ambient work place air. It is measured through 3 the use of air sampling equipment and expressed uCi/ml (Bq/m ). 3.2.1.6. Contamination: This is a measure of the amount of uranium surface I contamination in the work environment, expressed in units of 2 dpm/100cm. 3.2.2. Reduction in Effluent: In the interest oflimiting exposures to the public, efforts will be made to reduce effluent releases containing radioactive contaminants. These efforts will include, but not be limited to, recapturing decontamination solutions for recycling (cleaning and reuse), and maintaining ventilation controls in areas where radioactive materials are being handled. 3.2.3. Limiting Intakes By Use of Engineering Controls: The largest potential contribution to total effective dose equivalent for decommissioning workers is radioactive material inhaled during performance of theirjobs. Limiting intake l will effect a direct reduction in total effective dose equivalent. The preferred method oflimiting intake is the use of engineering controls. The primary engineering control that will be used during the project is ventilation control. For tasks with the potential to generate airbome radioactive material, efforts will be made to minimize the spread of such material by isolating the work area (through use of glove bags, containment tents, or equivalent) and I providing ventilation through a 11 EPA filtration unit. In cases where engineering controls are not adequate to protect the workers, respiratory protection may be used in accordance with a respiratory protection program. 3.3. Health Physics Program: 3.3.1. ALARA policy: It is the policy of Combustion Engineering to conduct its business in a manner which ensures that its facilities are in compliance with I radiation protection and other applicable regulations and that the operation of these facilities will not be detrimental to the environment. In implementing this policy with regard to decontamination and decommissioning, Combustion Engineering will ensure that radiation exposure to personnel is maintained As Low As Reasonably Achievable (ALARA). For activities carried out within the scope of this plan, responsibility for establishing and ensuring adherence to this policy shall rest with the Program Manager, Decontamination and Initial Issue 3-5 3/30/93 I I

i ) I Decommissioning. The policy shall be implemented through appropriate delegation to the Project Manager, Radiation Safety Officer, and other applicable line managers. The ALARA program that will be in place during the project is described in section 3.2 above. 3.3.2. Radiation work permit procedures: Work with contaminated materials will be I covered by procedures and/or radiation work permits. The Radiation Safety OfHeer will approve the issue of radiation work permits. Radiation protection and industrial safety technicians will be responsible for ensuring the proper I implementation of radiation work permits. Each radiation work permit will be reviewed at a frequency established by the Radiation Safety OfHeer as part of the radiation work permit approval process and will be updated as appropriate. Radiation work permits will be retained, as a minimum, for a period of three years or until the license has been terminated. 3.3.3. Personnel contamination control: Radiologically Controlled Areas for purposes of protection of individuals from exposure to radiation and radioactive materials will be established. Radioactive material, airborne radioactivity, radiation, and contaminated areas will be identified and conspicuously posted. Radiological protection measures required in those areas will be specified in decommissioning procedures, radiological protection instructions, and/or radiation work permits. I Protective clothing will be required for entry into contaminated areas. The type (s) of protective clothing required will be commensurate with the levels and type (s) of contaminants present and the nature of the work assignment and will be specified in procedures, the radiation work permit, or a radiological protection instruction. Personnel exiting contaminated areas will be required to survey themselves after removing protective clothing to ensure that they are free of contamination. I Emergency evacuations will be an exception to the personnel survey requirement. Personnel survey instruments will be located at the normal exits from contaminated areas and personnel will be trained in their use. 3.3 A. Ventilation: Ventilation systems are designed and maintained to limit the spread of airborne contamination outside the pellet shop by drawing air from the pellet shop and exhausting it through the IIEPA filtration units to outside the pellet shop. This process maintains lower air pressure in the pellet shop than in surrounding areas.11 EPA ventilation systems will be sampled using I guidance provided in ANSI N13.1 - 1969, " Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities". The adequacy of air effluent control will be determined by representative stack sampling to demonstrate Initial Issue 3-6 3/30/93 I I

P compliance with applicable regulations. Sampling will be performed continuously when the IIEPA ventilation systems are in operation. Samples will be collected, analyzed, and documented at a minimum of once per week I during ventilation system operation. Ventilation systems senicing laboratory type hoods and/or other enclosures I where uncontained radioactive material is handled will be maintained during working operations at an average face velocity of at least 100 linear feet per minute (100 LFPM). Face velocity measurements will be obtained and I documented weekly when the hoods are in use. I Ventilation air from the fixed exhaust air systems will be passed through HEPA filters or a combination of pre-filters and IIEPA filters. Ventilation systems which contain HEPA filter installations will be equipped with continuous pressure differentia! measuring / indicating systems which will be read and the readings recorded weekly during operation. Thr maximum differential pressure permitted across the HEPA filter will be four inches (4") of water gauge or as specified by the system manufacturer. 3.3.5. Work area air sampling: During operations, airborne radioactivity will be sampled continuously in areas where operations are likely to result in an individual's intake of uranium exceeding ten percent (10%) of the Annual Limit on intake (AL1) as specified in appendix B to 10 CFR s 20.1001-20.2401, Table one, column two (as amended through 5-21-91). The ALI for class Y uranium is 4E-2 pCi (2E3 Bq). The air will be sampled for particulate r activity as required by radiation work permits. When required by the radiation I work permit, air will be sampled on a continuing basis when work is in progress using personal (lapel) air samplers and/or representative fixed position air samplers. The filter papers from these samplers will be changed and I analyzed at least once per work week and the results documented. Time will be allowed for the decay of short lived natural activity prior to analysis. Calcu ations of airborne radiot.ctivity concentration will include correction l factors for background, counting efficiency, and self absorption, if any. Counting equipment and techniques will provide for the detection of airborne radioactivity concentrations at or above 10% of Derived Air Concentration I (DAC) for uranium. The capability of fixed position air samplers to represent the breathing zone will be demonstrated before they are used to routinely assign intake. 3.3.6. Radioactivity measuring instruments: Instruments used for radiation detection and measurement shall have capabilities as follows (more than one instrument may be utilized to cover the specified range): l I Initial Issue 3-7 3/30/93 I I

Alpha / Beta Counting Systems: 1 DPM to IE7 DPM (disintegrations per minute) Alpha Survey Meters: O CPM to 5E5 CPM (counts per minute) Beta / Gamma Survey Meters: 0.1 mrad /h to 500 mrad /h (millirad per hour) Gamma Survey Meters: 0.001 mR/h to 5 mlUh (millirem per hour) Minimum detectable actisity is calculated, at a 95% con 6dence, according to the following formula: I C# MDA = K 2.71 + 4.65 ) Ti > where: I minimum detectable activity level in MDA = 2 disintegrations / minute /100cm background counts C,, = background count time T,, = other factors such as probe area, detector K = I ef6ciency, sample size, chemical yield, and number of emissions of radiation being measured per disintegration of the radionuclide, that may be applied to correct for the specific survey and/or analysis being performed Radiation detection instruments will be calibrated semi-annually and/or follow' g repairs which may affect the accuracy of the instrument. Alpha / Beta m counting systems will be checked daily to verify that background and ef6ciency I remain within specified operating parameters. I Records of instrument calibrations will be retained, as a minimum, for a period of three years. 3.3.7. Radiation exposures: Dosimeters capable of detecting and measuring beta, gamma, and x-ray radiation will be supplied to personnel who are likely to receive an exposure of ten percent (10%) of the limits specified by 10 CFR s 20.1201 (as amended through 5-21-91). These limits are as follows: Initial Issue 3-8 3/30/93 I

3.3.7.1. Total effective dose equivalent equal to five (5) rems. 3.3.7.2. The sum of the deep dose equivalent and the committed effective I dose equivalent to any individual organ or tissue other than the lens of the eye being equal to fifty (50) rems. 3.3.7.3. An eye dose equivalent of fifteen (15) rems. 3.3.7.4. A shallow dose equivalent of 50 rems to the skin or to any extremity. An investigation of personnel radiation exposures will be conducted if credible dosimetry results exceed eighty percent (80%) of applicable limits specified by 10 CFR { 20.1201 (as amended through 5-21-91). In accordance with the provisions of 10 CFR s 20.2106 (as amended through 5-21-91), records of personnel radiation exposures will be retained, as a minimum, until the license has been terminated. I 3.3.8. Surface contamination: Contamination surveys will be performed on a routine basis to evaluate the potential spread of radioactive contamination. Routine contamination surveys will be performed in the pellet shop at a minimum of once per week. Routine contamination surveys will be performed in the balance of the building 17/21 complex once per calendar quarter. Survey frequency may be increased as appropriate during decontamination work. Surveys conducted in support of work performed under a radiation work permit may be used to meet the weekly survey requirement. Decontamination will be performed by the next regularly scheduled working shift on areas that are 2 discovered to contain loose contamination greater than 5,000 dpm!!00cm, Records of contamination surveys will be retained, as a minimum, for a period of two years. 3.3.9. Determination ofInternal Exposures: Individuals who are likely to receive, in 1 year, an intake in excess of 10 percent of the Annual Limit on Intake (ALI) shall be monitored for exposures to radioactive material. For individuals likely to receive an intake in excess of 10 percent of the All, internal dose will be assessed by taking suitable and timely measurements of: 3.3.9.1. Concentrations of radioactive materials in the air in the work area; or 3.3.9.2. Quantities of radionuclides in the body; or 3.3.9.3. Quantities of radionuclides excreted from the body; or Initial Issue 3-9 3/30/93 I I

3.3.9A. Any combination of the measurements specified in 3.3.9.1 through 3.3.9.3 above. In accordance with the protisions of 10 CFR f 20.2106 (as amended through 5-21-91), records of estimated intake or body burden of radionuclides will be retained, as a minimum, until the license has been terminated. 3.3.10. Environmental monitoring: With the cessation of uranium processing operations, the potential for a release of radioactive material to the environment is greatly diminished. As such, environmental monitoring will consist of continuous sampling of the air discharged from the pellet shop of building 17 through the llEPA filter systems. The air will be isokinetically sampled whenever tY IIEPA systems are in operation. Liquid effluents will also be sampled prior to discharge to demonstrate that the average activity concentrations are equal to or less than 3E-7 pCi/ml gross alpha activity. Upon completion of decontamination and decommissioning activities, environmental monitoring will cease. Records of environmental monitoring data will be retained, as a minimum, for a period of two years. I 3.3.11. Internal Inspections: The inspection function is a normal part of the radiological protectien iob function. On a monthly basis, radiological protection technicians shall perform a documented inspection using a prepared i checklist to review the conduct of facility operations. Any time the technicians find discrepancies, the cognizant supersisor will be informed of the remedial actions to be taken and a written deficiency report will be turned over to the Radiation Safety Officer (RSO). Completed inspection checklists shall be signed by the technician performing the inspection and turned over to the RSO. Records of Internal Inspections will be retained, as a minimum, for a period of three years from the date of issue or until the closure of inspection findings, whichever is longer, or until the project is completed. 3.3.12. Audits: Audits will be performed to verify that operations are being conducted according to established criteria. Audits will be conducted in accordance with a -I written plan. I Individual (s) conducting audits will meet the qualification requirements of the Radiation Safety Officer. Reports of audit findings will be submitted to the Project Manager for disposition of the findings. (Copies of audit findings will be provided to the Program Manager.) Audits of the decommissioning project Initial Issue 3-10 3/30/93 I l

I will be conducted at least annually. As a minimum, the audits will review ALARA trends and effluent releases. Records of audit reports shall be retained, as a minimum, for a period of three (3) years from the date ofissue or until the closure of audit findings, whichever is longer, or until the project is completed. 3.3.13. Safety Committee: The Safety Committee will be comprised of, as a minimum, the Program Manager (chair), Radiation Safety Officer, Project Manager, and Licensing. The Safety Committee will review abnormal occurrences and significant decommissioning actions. As a minimum, the Safety Committee will meet quarterly to discuss safety practices and trends, including ALARA. Records of Safety Committee proceedings will be retained, as a minimum, for a period of three (3) years from the date ofissue or until closure of Safety Committee findings, whichever is longer, or until the project is completed. 3.3.14. Investigations and Reporting: Abnormal occurrences will be investigated in accordance with written procedures. Reports to the Nuclear Regulatory Commission will be made in accordance with the applicable Federal Regulations. The level ofinvestigation and the need for corrective action will be determined based on the severity of the incident. The Safety Committee may be charged with the responsibility for investigating abnormal occurrences and recommending corrective action (s), as appropriate. Records ofinvestigations of abnormal occurrences reported to the Nuclear Regulatory Commission will be retained, as a minimum, for a period of three (3) years after closure of the investigation or until the project has been completed. } 3.3.15. Records: Records pertaining to health and safety, abnormal occurrences, inspections, audits, instrument calibrations, employee training, personnel exposures, Radiation Work Permits, and environmental monitoring data will be retained for the periods specified in this plan or in the governing regulations, whichever is longer. Records of the final status survey will be maintained demonstrating that the I j building 17/21 complex is acceptable for unrestricted use, including but not limited to the following: i 3.3.15.1. Final status survey records: Radiation and contamination survey results will be recorded on standardized forms similar to the form depicted in figure 3-1. Information that will be recorded on the final Initial Issue 3-11 3/30/93 1

status survey form includes: the name of the individual performing the survey; the date of the survey; the serial numbers, calibration due dates, backgrounds, and minimum detectable activities of the I instruments used for the survey; and the counting uncertainty associated with each survey measurement. Specific surveys will contain radiological data reported in standard units as follows: Gamma radiation surveys will show data in units of pR/h (microRoentgens per hour). 2 Direct alpha surveys will show data in units of dpm/100cm (disintegrations per minute per 100 square centimeters). 2 Alpha' Beta scans will show data in units of dpm/100cm (disintegrations per minute per 100 square centimeters). Soil and other volumetric sample reports will show data in units of pCilg (picoCuries per gram) of "U,2"U and *'U. 2 3.3.15.2. The final status survey records referenced in 3.3.15.1 above will be retained, as a minimum, for the longer of a period of three years following the date the record is made or until the license has been terminated. These records will be considered adequate to demonstrate and permanently document that the building 17/21 complex has been free released and is available for unrestricted use. I I I I I i Initial Issue 3-12 3/30/93

I FIGURE 3-1 SAMPLE RADIOLOGICAL SURVEY FORM DIRECT SURFACE ACTIVITY SHEET ! Sample # Dpm\\100ctr9 4\\-1.96 o i Sample # Dpm\\1000rr# +\\- 1.96 o I i 4 i l l l -i [ l i 7 I i t l I l l l 4 l i i i I 3 J } 7_ f 4 { i i j _.L l i l i q l___ _j L { l i 1 4 1 t 1 _4 I 1 j i I i l i J l l u 1 i 1 l l i } i i 4 -.1 I \\ \\ \\ I I s ! INSTRUMENTATION COUNTING DATA SURVEY BY l 9 jINSTRUMENT SERIAL CAL i BKGD EFF l MDA PRINT NAME: NO. DUE l (cpm) I (dpm) ~ SIGNATURE: I l l i al DATEmME: l l i _. _. _ _ 1 I il r tj Initial Issue 3-13 3/30/93

3.4. Contractor Personnel: Contractor personnel will be protected under the same procedures as Combustion Engineering personnel. Implementation of those procedures I is described in sections 2.5 and 3.3 of this plan. Section 2.9 provides further information on contractor personnel that may be used during decontamination and decommissioning. I 3.5. Waste Management: 3.5.1. Radioactive Waste: Upon shutdown of uranium bearing manufacturing operations, much of the process equipment located in the pellet shop will be removed from the building 17/21 complex and sent to another licensed facility. I Radioactive waste generated after that time will be from decommissioning activities. Radioactive wastes anticipated from decommissioning are described below. 3.5.1.1. It is estimated that the volume of radioactive waste generated from decommissioning will be approximately two thousand (2,000) cubic feet. 3.5.1.2. Since the revitalization project (see section 3.1.2), contamination levels in the pellet shop have remained low. As a result, radioactive i waste generated during decom.missioning should contain only low levels of contamination. 3.5.1.3. Radioactive wastes from decommissioning should consist of a combination of the following materials: concrete, insulation, structural materials, decontamination supplies, and floor coverings. 3.5.1.4. Due to the low specific activity and low concentration of uranium present, as well as the low abundance of penetrating radiation from the radionuclides of concern, radioactive waste from decommissioning are not expected to generate significant exposure rates. As such, temporary storage of these wastes may be performed according to the provisions of Appendix B. 3.5.2. Mixed / Hazardous Wastes: There is currently no mixed waste generated in buildings 17 and 21. Administrative controls will be in place to ensure that mixed waste is not generated during the project. I 3.5.2.1. Prior to their use, decontamination methodologies will be evaluated to ensure that they meet the requirement stated in section 3.5.2 above. 1 I Initial Issue 3 14 3/30/93 I I

1 I 3.5.2.2. When the Resource Conservation and Recovery Act (RCRA) storage I areas are closed, samples will be appropriately analyzed for both radioactive and chemical content. 3.5.2.3. Materials listed as hazardous in 40 CFR 261 will not be used m processes which present a potential of their coming into contact with j I radioactive materials. As the majority of work to be performed during the D&D process will involve radioactive material, it is not I anticipated that work with hazardous chemicals will occur in i conjunction with this project. I 'I I I l ? 1 I I I i l l I i l l t Initial Issue 3-15 3/30/93 I l L

l 3 P l I i I , I SECTION 4 i-l I I I PLANNED FINAL STATUS SURVEY I

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Initial Issue 4-1 3/30/93 I-1

4.0. Planned Final Status Suney: Upon completion of the decontamination activities described in this plan, areas within the scope of the project will be surveyed to determine if they meet the criteria for release as specified in Table 4-1, 4.1. Final Status Survey Instrumentation: Instrumentation used for the final status survey will be selected as appropriate for the radiations the survey seeks to identify. Detailed instructions regarding instrument selection, calibration, use, and control will be included in final status survey procedures. Surveys will be performed using one or more of the following instruments and detectors: 4.1.1. Survey instruments such as the Eberline ESP-2 or similar. The ESP-2 is a microcomputer based survey meter with data logging and computer download capabilities. It can be used with a variety of different probes including scintillation, Geiger-Meuller, and proportional, giving it the capability to be I used to detect alpha, beta, and/or gamma radiations. The ESP-2 can be operated in either ratemeter or scaler modes and has both a digital meter and an audible count rate indicator. Its data logging capabilities make it useful for a final status survey. 4.1.2. Gamma survey instruments such as the Ludlum 19 micro R meter or similar. The micro R meter is useful for niaking low level gamma measurements. 4.1.3. Radiation monitors such as the Eberline RM-20 or similar. The RM-20 is a compact, alarming, count rate meter with variable high voltage. It can use a variety of detectors, making it useful for both beta-gamma and alpha monitoring. 4.1.4. Gas flow proportional detectors such as the Eberline llP-100A or similar. The IIP-100A is a hand held probe with a 100 square centimeter active probe area. I lt is capable of detecting both alpha and beta-gamma radiations and has a 4n cfficiency of approximately 20% - 40% depending on the type of radiation being measured. Gas flow proportional probes may be set up to detect alpha particles only, beta / gamma radiations only, or a combination of alpha and beta / gamma. 4.1.5. Alpha scintillation probes such as the Eberline AC-3 or similar. Used for direct alpha surveys, the AC-3 has an active zine sulfide detector area of 50 I square centimeters, and 4n alpha efficiency of approximately 14%. Alpha scintillators are useful for personnel and equipment surveys. I 4.1.6. Geiger-Meuller probes such as the Eberline llP-210 or similar. The llP-210 is a thin window GM tube that is useful in detecting low energy beta particles. I Initial Issue 42 3/30/93

I TABLE 4-1 DECOMMISSIONING RELEASE CRITERIA Following are the release criteria to which areas within the scope of the project will be decontaminated. Ensuring that contamination is at or below these levels is sufficient to demonstrate that the health and safety of the public and the environment is protected and that no use of other risk assessment methodology shall be required. Meeting these criteria will I demonstrate that the included areas have been successfully decontaminated and are available for unrestricted use. The criteria are as follows: I Catecorv Limit 2 Removable surface contamination 51,000 dpm a/100 cm above background 2 Average Fixed Surface Contamination 5 5,000 dpm n/100 cm above background I 2 Maximum Fixed Surface Contamination 515,000 dpm a/100 cm above background Average Residual Soil Contamination ' 5 30 pCi/ gram above background Exposure Rate at One Meter From Surfaces 5 5 pR/h above background .) His limit will also apply to other materials discovered to be volumetrically contaminated. i Initial Issue 43 3/30/93 I

The 4n efficiency for this probe is approximately 10% - 20%. GM probes are useful for personnel and equipment monitoring surveys. 4.1.7. Gamma scintillation probes such as the Eberline SPA-9 or similar. The SPA-9 is a 2 inch by 0.5 inch sodium iodide detector designed for detection of low to medium energy gamma radiations. It's 4x efficiency is approximately _3.5%. I-Gamma scintillators are useful for open land scanning surveys. 4.1.8. Gamma spectroscopy on samples requiring analysis will be perfomied using a I multichannel analyzer such as the Canberra Series 80 4096 channel multichannel analyzer and a 30% intrinsic germanium solid state detector or similar. 4.2. Methodology for Ensuring Adequate Survey Coverage: This section details the final status survey methodology. The survey is designed to ensure that systems, equipment, structures, and land areas applicable to the project are included and that the data generated are statistically meaningful. The following survey techniques are specified in the methodology and will be performed as described here. Scans for alpha and beta radiations: These scans involve moving a detector such as an HP-100A, AC-3, or similar over the surface to be surveyed at a rate approximately equal to one detector width per second. The monitoring instrument will be configured to detect both alpha and beta emissions and to detect count rate increases over background. Direct alpha measurements: Direct alpha measurements will be made by placing a I detector such as the HP-100A or similar within one centimeter of the measurement location and holding it there for a specified length of time. The monitoring instrument will be configured to detect alpha radiation and to integrate the counts detected over the total measurement time. Actisity will be derived by dividing the total counts by the measurement time. Gamma scans: Gamma scans will be performed by holding a detector such as an SPA-9 or similar as close to the ground as practical and moving it from side to side while walking over the surface at approximately one half of one meter per second. The I monitoring instrument will be configured to detect gamma radiation and to detect count rate increases over background. L Gamma exposure rate measurements: Gamma exposure rate measurements will be made by holding an exposure rate meter such as the Ludlum 19 (pR meter) at one meter over the surface and reading the exposure rate from the meter face. I initial Issue 44 3/30/93 l I-

I Buildine 17 Pellet Shoo 4.2.1. Suneys of interior building surfaces will be performed based on a grid system I of measurement locations. Grids for floor md wall surfaces will be set up at one meter intervals, will cover the pelbt shop floor and wall surfaces and will be clearly marked on the surfaces to be surveyed. A system of grid designations will be developed such that positions on any gridded surface are readily identifiable. Ceiling joists and moldings will be used to delineate grid blocks for the pellet shop ceiling surfaces. A system of grid designations will also be developed for the ceiling grid blocks such that any ceiling surface can be readily identified. 4.2.2. Scans for alpha and beta radiations will be performed on interior building surfaces to detect the presence of areas with elevated contamination levels. 4.2.3. Direct alpha measurements will be made of interior building surfaces at fixed I points on the grid system. Measurements will be made at 2 meter intervals on floor and wall surfaces (Figure 4-1). The measurement interval specified below for the pellet shop ceiling has been chosen to take advantage of the natural demarcations represented by the ceiling joists and panel molding strips. The joist / molding interval for the ceiling of the l pellet shop (Figure 2.3) is 1.2 meters on a north / south axis and 0.8 meters on an east / west axis. Thus, the area of each " grid block" is 0.96 square meters. The joist / molding interval in the pellet shop annex (Figure 2-3) is 1.5 meters on a north / south axis and 0.6 meters on an east / west axis, yielding a " grid block" area of 0.9 meters. Measurement intervals for the pellet shop ceiling will be 2.4 meters on the north / south axis and 1.6 meters on the east / west axis. Measurement intervals for the pellet shop annex ceiling will be 3.0 meters on the north / south axis and 1.2 meters on the east / west axis. (See Figure 4.2 for details.) I 4.2.3.1. If conditions at the time of final survey make detection oflevels 2 <l.250 dpm n/100cm impossible during scanning measurements, direct measurements will be made at 1 meter intervals on floor and wall surfaces,1.2 meter (north / south) and 0.8 meter (east / west) intervals on the pellet shop ceiling surfaces, and 1.5 meter (north / south) and 0.6 meter (east / west) intervals on the pellet shop annex ceiling surfaces. (See Figure 4.2 for details.) 4.2.3.2. Enough direct measurement locations will be evaluated to assure that i the confidence level of the survey is at least 95%. (See section 4.3 1 lI 1 Initial Issue 45 3/30/93 i lI 1 i

l Figure 4-1 Standard Measurement / Sampling Pattern For l Systematic Grid Survey of Structure Surfaces O A O A O s v a v a I l C) () u t) u Q, I <m l ('Y () V7 () VY N/ I <m l (} k) N) N) N) NJ I a em r.a em ra r V O V O g I g MEASUREMENT LOCATIONS IF SCANNING TECHNIQUE IS CAPABLE TO DETECTING < 1,250 DPM u i I O IS NOT CAPABLE OF DETECTING < 1,250 DPM MEASUREMENT LOCATIONS IF SCANNING TECHNIOUE a I N O 3 ~M 0 I METERS I initial lssue 4-6 3/30/93

g Figure 4-2 Grid Demarcation of Pellet Shop Ceiling PFI I FT SHOP CEfLING (EXCLUDING ANNEX) n-1 m n n o I f JOlST (12cmwdh) ) l . i{ a .i .1 ee - @ O,

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O O: LO- "C C=Q ~] ? .] l S MATERLAQ l l [l iI }i-I' l!- ip .4 a o a w g j MOEDING DTRIP$ (3.6 cm wd:h) (Not to Scale) ANNEX CEILING _._2__ _I I E E E .m .. __ g JO:ST (2 x 12 cm WIDTH) _[ e. = GRID AFIEA 9

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5_, .~} ~b u 1 O r i CE! LNG ~ g SURFACE f f ~, f ~, f f ~, f l + g (sTEEu L L > t > L > L > t ; i r j I UNES DEPICT b l omsiow armtEn 4 U O ' :1 9 h ) 0 0 i b4 l I PANELS 60 cm ~ ~ ~ I 9 MEASUREhENT LOCATIONS IF SCANNING TECHNIQUE IS CAPABLE OF I DETECTING s 1,250 DPM u "^S""'"'*^"*S'"S'^**'" ' ' ' " * ' "' 'S * ' O "APABLE OF DETECTING s 1,250 DPM a i a E C l Initial Issue 47 3/30/93

for a description of the methods used for data analysis and comparison with guideline values.) 4.2.4. Gamma exposure rate measurements will be made at one meter from the floor and wall surfaces at a minimum of one per ten (10) square meters to a distance of two (2) meters above floor level. 4.2.5. Removable contamination measurements will be made at each location where a direct measurement is made. 4.2.6. Paint samples will be taken from walls where paint remains following completion of remediation activities. Enough paint samples will be taken to i ensure at a 95% confidence level that the surface contamination level meets the criteria specified in Table 4-1. (See section 4.3 for a description of the methods used for data analysis and comparison with guideline values.) Buildine 17. Clean Shon. Clean Side Mezzanines. Office Areas: and Buildine 21 Interior: These have been maintained as clean areas and have low potential for j contamination from past commercial operations. These areas will be surveyed as follows: 4.2.7. Surveys of interior building surfaces will be performed based on a grid system of measurement locations. Grids will be set up at one meter intervals and will be clearly marked on the surfaces to be surveyed. A system of grid designations will be developed such that positions on any gridded surface are readily identifiable. I 4.2.7.1. Grids will cover the floor areas and interior wall surfaces to a distance of two meters above floor level. 4.2.8. Scans for alpha and beta radiations will be performed on gridded surface areas to detect the presence of areas with elevated contamination levels. 4.2.9. A minimum of 30 direct alpha measurements will be made on gridded interior building surfaces at fixed points on the grid system at the rate of approximately one per fifty square meters. 4.2.9.1. A minimum of 30 direct alpha measurements will be made on non-I gridded areas at the rate of approximately one per fifty square meters. 1 4.2.9.2. Enough direct measurement locations will be evaluated in the non-l gridded and gridded areas to assure that the confidence level of the survey is at least 95%. (See section 4.3 for a description of the I i l Initial Issue 48 3/30/93 1 I i

t methods used for data analysis and comparison with guideline values.) 4.2.10. At each location of a direct alpha measurement, a gamma exposure rate measurement will be made at one meter from the building surface. 4.2.11. Removable contamination measurements will be made at each location where a direct measurement is made. Buildine 17. Roof: The roof of building 17 measures approximately 120 feet by 300 feet, with a total area of approximately 36,000 square feet. It consists of aggregate I layers comprised of fiber board, tar paper, roofmg tar, and gravel approximately six inches thick, giving the roof a volume of approximately 18,000 cubic feet. The roof was subject to gravitational settling and atmospheric deposition of contamination from the exhaust of the llEPA ventilation systems. The roof will be surveyed as follows 4.2.12. Surveys of roof surfaces will be perfomied based on a grid system of measurement locations. Grids will be set up at one meter intervals. will cover the roof surfaces and will be clearly marked on the surfaces to be surveyed. A system of grid designations will be developed such that positions on any } gridded surface are readily identifiable. 4.2.13. Scans for alpha and beta radiations will be performed on roof surface areas to detect the presence of areas with elevated contamination levels. 4.2.14. Direct alpha measurements will be made of roof surfaces at fixed points on the grid system. Measurements will be made at two meter intervals (Figure 4-1). t 4.2.14.1. If conditions at the time of final survey make detection oflevels I <1,250 dpm n/100cm impossible during scanning measurements, i 2 direct measurements will be made at I meter intervals. 4.2.14.2. Enough direct measurement locations will be evaluated to assure that the confidence level of:he survey is at least 95%. (See section 4.3 for a description of the methods used for data analysis and I a comparison with guideline values.) y t 4.2.15. Gamma exposure rate measurement will be made at one meter from the roof i surface at a minimum of one per ten (10) square meters. l [ 4.2.16. Removable contamination measurements will be made at each location where a direct measurement is made. I Initial Issue 49 3/30/93 i i i

L I Buildines 17 and 21. Exterior Walls. and Buildine 21 Roof: 'lle lik.clihood of contamination on these surfaces is considered low. The exterior surfaces of buildings 17 and 21 will be surveyed as follows. I 4.2.17. Scans for alpha and beta radiations will be performed on a minimum of 10% of exterior wall and roof surfaces. 4.2.18. A minimum of thirty direct alpha measurements will be made on each exterior wall at biased locations at the rate of approximately one per fifty square meters. Measurement locations will be selected at points considered to have a greater likelihood of being contaminated than others. j I 4.2.18.1. Enough direct measurement locations will be evaluated to assure that the confidence level of the survey is at least 95%. (See section 4.3 I for a description of the methods used for data analysis and comparison with guideline values.) 4.2.19. At each location of a direct alpha measurement, a gamma exposure rate measurement will be made at one meter from the building surface. l 4.2.20. Removable contamination measurements will be made at each location where a direct measurement is made. Open 1,and Areas Inside and Within One Meter Outside the Buildine 17/21 Complex Fence Line: 4.2.21. The land areas will be gridded at ten meter intervals and will receive gamma scanning coverage. 4.2.22. Soil samples will be obtained from the top 15 centimeters of soil at four locations in each grid located approximately equidistant from the four grid block corners and the grid block center (Figure 4-3). 4 2.22.1. If sample results in any grid block are greater than 3 times the 30 pCi/g guideline value specified in Table 4-1, soil samples from I additional points, located along the ten meter grid lines, at the block corners and midway between the block corners, as well as at the block center, will be obtained from the grid in question (Figure 4-3). l 4.2.22.2. Enough sample locations will be evaluated to assure that the I confidence level of the survey is at least 95%. (See section 4.3 for a description of the methods used for data analysis and comparison with guideline values.) l Initial Issue 4 10 3/30/93 I

I l Figure 4-3 Soil Sampling Locations g I 10 M

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O O g i I I I I 1 I a a g e O o o I ~ I e e i I 2-0 0 o t g g Initial Systematic Sampling Locations i g O Additionai Sampling Locations (required in a given grid j block if any original samples in that block are > 30 pCi!g) ] Initial Issue 4 11 3/30/93

I 4.2.23. Gamma exposure rate measurements will be made one meter from the land surface at the location of each of the original four soil samples in each grid block. Drainaee Swales West and South of the Buildine 17/21 Complex Fence: 4.2.24. Thirty soil samples will be taken from each swale at random locations. Results of these samples will be compared with applicable guideline value (Table 4-1) in accordance with the methods specified in sections 4.3.4 and 4.3.5. 4.1.5. After any remediation indicated by the thirty initial samples is completed, the need for additional samples will be determined by the method specified in section 4.3.6. I 4.2.26. Additional samples taken will be compared with the applicable guideline value as specified in sections 4.3.4 and 4.3.5. Paved Areas Within the Buildine 17/21 Complex 4.2.27. Paved areas will be gridded at one meter intervals, with the grids covering the paved areas. The grids will be clearly marked on the surfaces to be surveyed. A system of grid designations will be developed such that positions on any gridded surface are readily identifiable. I 4.2.28. Scans for alpha and beta radiations will be performed on a minimum of 10% of paved areas to detect the presence of areas with elevated contamination levels. I 4.2.29. A minimum of 30 direct alpha measurements will be made on paved area surfaces at the rate of approximately one per fifty square meters. 4.2.29.1. Enough direct measurement locations will be evaluated to assure that the confidence level of the survey is at least 95%. (See section 4.3 I for a description of the methods used for data analysis and comparison with guideline values.) 4.2.30. Gamma exposure rate measurements will be made at one meter from the paved surface at a minimum of one per ten (10) square meters. 4.2.31. Removable contamination measurements will be made at each location where a direct measurement is made. I I Initial Issue 4-12 3/30/93 I I

s Miscellaneous Fouinment. Materials. and Other Non-Permanent items Within the Iluildine 17/21 Complex Fence Line: Equipment, materials, and other non-permanent items not included in any other category listed under this section are not likely to be contaminated and will be surveyed as follows: 4.2.32. Equipment and materials will be scanned to detect areas with elevated I contamination levels. 4.2.33. Direct measurements will be made on miscellaneous equipment and/or materials I at representative points on accessible surfaces. 4.2.33.1. Enough direct measurement locations will be evaluated to assure that the confidence level of the survey is at least 95%. (See section 4.3 for a description of the methods used for data analysis and comparison with guideline values.) 4.3. Methodology for Data Analysis and Comparison with Guideline Values: The I following is a description of the methodology that will be used to analyze and test the data gathered during the final status survey and to compare that data to the applicable guidelines. (Note: Data reported during the final status survey will be recorded exactly as it is observed, even when the values are less than the MDA calculated for the survey instrument. This will avoid biasing the data and affecting the statistical data analysis described in this section.) Elevated Areas of Activity pp p 4.3.1. Areas of elevated activity on buildings, structures, and paved areas: The limit for surface activity is three times the guideline value (Table 4-1) when 2 averaged over 100 cm. Activity over three times the guideline will require I additional remediation. Areas of elevated activity between one and three times the guideline value will be tested as follows to assure that the weighted average surf e activity level within the contiguous one square meter grid containing E _g the elevated area (s) is less than the guideline: r 4.3.1.1. Additional measurements v/ill be made to determine the activity and [ surface area of the elevated area. The weighted average of the one square meter grid will then be calculated, taking into consideration I the relative fraction of the grid that is occupied by the elevated area (s). The formula for calculating the weighted average is: n, n, n, R = - [ x, 1 -[ A, + [ y, A,

• i=1

- k=1 k=1 I Initial issue 4-13 3/30/93 f

where: 7= weighted mean including elevated area (s) systematic and random measurements at point i x = i n, number of systematic and random measurements = y, elevated area activity in area k = 2 fraction of one m occupied by elevated area k A, = n, number of elevated areas = 4.3.2. Areas of elevated activity in soil: The limit for residual soil radioactivity at i any location is three times the guideline value as shown in Table 4-1. Areas with elevated activity exceeding this level will be remediated. Areas with elevated activity between one and three times the guideline limit will be tested to assure that the average concentration is less than (100/A)"2 times the I' guideline value, where A is the area of the elevated activity in square meters. Areas with activity exceeding this limit will be remediated. If this condition is 2 satisfied, the weighted average activity in the 100 m contiguous area containing the region of elevated activity will be determined to assure it is within the guideline value. The formula for calculating this average is: n n; ng f, = 1 {, x, 1 -{ A, + { y,A, "" i=1 - k=1 - k=1 7, = weighted mean including elevated area (s) x, systematic and random measurements at point i = n, number of systematic and random measurements = I y, elevated area activity in area k = 2 fraction of 100 m occupied by elevated area k A, = n, number of elevated areas = General Area Measurements 4.3.3. Exposure rates: Exposure rates (in pR/h) will be compared directly with the guideline value. If the maximum exposure rate is greater than two times the guideline value over background, the area will be remediated and resurveyed. Comriarisons with Guidelines 4.3.4. Calculating average values: Guideline values (Table 4-1) established for surface activity, soil activity, and exposure rates are for average values, above background. Tc allow comparison of the survey data gathered during the final Initial Issue 4 14 '3/30/93 I

I L status survey with those guidelines values, the mean (x) value of the measurements in each measurement category will be calculated using the formula: I. n I" It ' i=1 where: average of all measurements - x = number of measurements made I = n, x, measured level at point i = I 4.3.5. Comparing survey data with guideline values: Average levels, calculated by following the methods in 4.3.4 above, will be compared to the guideline values shown in Table 4-1. If the averages exceed the applicable guideline values, the areas involved will receive further remediation and will be resurveyed. After the averages satisfy the guidelines, the results will be further evaluated to determine whether the data from each survey prosides a 95% confidence level l that the true mean activity level satisfies the guidelines. This test will be performed by calculating the average (section 4.3.4 above) and standard deviation of the data for a particular radiological parameter from each sun'ey l using all applicable measurement locations. The calculation for the sample standard deviation is: I n { (T-x)2 j i=1 h n-1 2 where: average activity (as calculated in section 4.3.4) 'g x = standard deviation E s, = the number of measurements made n = x, measurement at point i = If there are areas of elevated activity in the survey area being considered, the 2 weighted mean (as calculated in sections 4.3.1 and 4.3.2) of each one m of building surface, or one hundred m of land. containing an elevated area will be used as one of the x's in the calculations specified in sections 4.3.4 and 4.3.5 above. I Initial issue 4-15 3/30/93 +

Once the average and standard deviations have been calculated, the data will be tested for the desired 95% confidence level, relative to the applicable guideline value, using the following equation: I I s, Va " #I }n l-a, df v where: the calculated mean (from section 4.3.4) x = I decision value = 1, the 95% confidence level obtained from the = standard "t" distribution chart (Table 4-2): df I (degrees of freedom) is n-l. a is the false positive probability, that is the probability that is less than the guideline value if the true mean activity level is equal to the guideline value s, the standard deviation (section 4.3.5) = the number ofindividual data points used to = n determine x and s, The value of p, will be compared to the guideline value; if p, is lese than or I equal to the guideline, the area being tested will be considered to have met the guideline at a 95% confidence level. I I I I I I I Initial Issue 4-16 3/30/93 I I

TABLE 4-2 Factors for Comparison of Survey Data with Guidelines and Determining Additional Data Needs I Degrees of Degrees of l Freedom

• t,3 4 t,,,33 Freedom "

tys.4 t,.7.3s t 1 6.314 12.706 18 1.734 2.101 2 2.920 4.303 19 1.729 2.093 3 2.353 3.182 20 1.725 2.086 4 2.132 2.776 21 1.721 2.080 5 2.015 2.571 22 1.717 2.074 6 1.943 2.447 23 1.714 2.069 7 1.895 2.365 24 1.711 2.064 8 1.860 2.306 25 1.708 2.060 9 1.833 2.262 26 1.706 2.056 10 1.812 2.228 27 1.703 2.052 11 1.796 2.201 28 1.701 2.048 12 1.782 2.179 29 1.699 2.045 13 1.771 2.160 30 1.697 2.042 14 1.761 2.145 40 1.684 2.021 15 1.753 2.131 60 1.671 2.000 j 16 1.746 2.120 120 1.658 1.980 17 1.740 2.110 400 1.649 1.966 infinite 1.645 1.960 " Degree of freedom is the number ofitems of data minus 1; for values of degrees of freedom not in table, interpolate between values listed. Initial issue 4 17 3/30/93 ?

4.3.6. Identifying additional measurement / sampling requirements: If p, as calculated in section 4.3.5 above, is greater than the applicable guideline, but x, as calculated in section 4.3.4 above, is less than or equal to the applicable guideline, x, as well as s,, as calculated in section 4.3.5 above, will be used with the following equation to determine the total number of data points required to demonstrate that the activity level satisfies the guideline at the I desired 95% confidence level. 2 I s* [2 _, + 2 _p]2 n, = 3 3 .C - x. a where; n, number of data points required = guideline value g; = mean value x = s, sample standard deviation = Standard normal variables: a is the false positive Zs, Z = g probability, that is the probability that p, is less than Ca if the true mean activity is equal to Co, and p is the false negative probability, that is the probability that p, is greater than Co if the true mean activity is equal to Ca.. For this project, a false positive probability, a, of five percent (5%) and a false negative probability, D, of ten percent (10%) will be used. Once n, has been calculated, the number of measurements previously made will be subtracted from n, and the required additional measurements will be made. The additional measurements will be spaced uniformly over the survey area and will be gathered using the same methodology as that used for gathering the I initial data. 1he additional data will be combined with that previously gathered l and the acceptance testing will be repeated using the methodologies specified in sections 4.3.4 and 4.3.5 above. If the value of p, again fails to be less than or equal to the guidelines at the desired 95% confidence level, additional remediation will be performed and the survey process repeated. I I Initial Issue 4 18 3/30/93 I

I 4 4.3.7. Counting uncertainty: Counting uncertainty will be expressed in the same units used for the measurement with which it is associated and will be calculated by the following formula:

• • • + 5 s =

r 2 \\ I.a* to s where; s, counting uncertainty = c.n counts due to sample + background = s sample count time ts4a = counts due to background = co background count time I, = I I 4 I I I I Initial lssue 4 19 3/30/93 II I l

lI l I I .I SECTION 5 I I I I PHYSICAL SECURITY PLAN AND SPECIAL NUCLEAR MATERIAL .I CONTROL AND ACCOUNTING PLAN 'l PROVISIONS IN PLACE DURING DECOMMISSIONING l I 9 'I i Initial lssue 51 3/30/93

5.0. Physical Security Plan and Special Nuclear Material Control and Accounting Plan Provisions In Place During Decommissioning: 5.1. Physical security: Combustion Engineering will follow the security plan as detailed in the document entitled " Physical Security Plan for the Protection of Special Nuclear Material of Low Strategic Significance Used Within Buildings 5/17/21", Revision 5, I dated January 19,1990, as well as any subsequent revisions made in accordance with the provisions of 10 CFR 70.32 (d) and (e) (as amended through 12-31-91). 5.2. Special nuclear material control and accounting: Combustion Engineering will follow Part I, chapters 1.0 through 9.0 of the present NRC approved ' document entitled I " Fundamental Nuclear Material Control Plan", dated November 17,1987, as well as any subsequent revisions made in accordance with the provisions of 10 CFR 70.32(c) (as amended through 12-31-91). I I I I i I i l Initial Issue 5-2 3/30/93 LI

1 i ) l SECTION 6 I a 1 i I l l l FUNDING i. J 6 I I i I i Initial Issue 6-1 3/30/93 i

l

6.0. Funding

A decommissioning funding plan for the Windsor site areas used for commercial production activities under SNM-1067 was submitted to the NRC on July 2,1992, under a cover letter entitled, " Decommissioning Funding Plan", from John F. Conant to John W.N. Hickey, letter I.D. # ML-92-035. That plan continues to be applicable to this project. I I l I I I I I I I Initial Issue 6-2 3/30/93

I I I I I I I l I APPENDICE.3 I I I I I I I I I t Initial Issue 3/30/93 I lI

J l I I i l I 3 i g Appendix A I Uranium Characteristics t 'I I I I s Initial Issue A.] 3/30/93 I 4

ll Appendix A Uranium Characteristics The uranium processed in Building 17 during its 25 year history was by license less than 5% by weight uranium 235. The great majority of this material was less than 4% by weight U-I 235. As discussed in Sections 1 and 3 of this plan, the uranium was received as either uranium dioxide powder (UO ) or uranium dioxide pellets. The chemical form of uranium is 2 class Y material, which is a relatively insoluble compound. Isotopic Ratios The interpretation of sample data requires knowledge of the isotopic composition of the material. When analyzing volumetric samples containing uranium by gamma spectroscopy, it I is important to know the ratio of the uranium isotopes of concern that are present. By knowing the uranium isotopic ratios, the radioactivity of the alpha only nuclides can be inferred by measuring the gamma emitting isotopes. Figure A-1 shows the uramum series radionuclides of concern with their short lived beta emitting progeny. During the enrichment of uranium the natural isotopic ratio is obviously changed. Since the principle radionuclide of concern. 23dU, is an alpha emitter with no significant gamma rays emitted, the amount of 23d radioactivity due to the U isotope must be inferred. An evaluation has been conducted to 23d establish U ratios so that the total uranium activity can be inferred based on the measurement of the other uranium and uranium progeny isotopes. I A U to "U radioactivity ratio of 23 I has been determined by analyzing over 230 23d 2 uranium samples by mass spectroscopy. The enrichments cf the samples ranged from 1.7% 2"U to 4.20% "U. Using the same set of data a "U to "U ratio of 30 : I has been 2 2 determined. The use of these ratios is considered acceptable for demonstrating compliance with contamination guidelines for volumetric samples such as soil that are analyzed by gamma I spectroscopy. Alpha to Beta Ratios The free release criteria as shown in Table 4-1 is given in disintegrations per minute per 100 2 cm (DPM) of alpha radioactivity. In some instances it may be useful to measure the beta I radiations from the progeny which are in secular equilibrium with the uranium parent. The assumption of secuhtr equilibrium of the beta emitting progeny is easily justified based on the long time period since the removal of the progeny in the conversion of UF to UO. The time I 2 elapsed since conversion has allowed the reestablishment of secular equilibrium of the short lived beta emitting progeny. Using the same set of mass spectroscopy data referenced above, an average activity ratio has 234 been determined in which 77% of the total uranium radioactivity is due to U,3% of the total uranium radioactivity is due to "U and 20% of the total uranium radioactivity is due to 2 Initial Issue A-2 3/30/93 I

{ i I 2"U. Ilased on these ratios, and the fact that one beta particle is emitted per alpha decay of 2"U, two beta particles per alpha decay of 2nU, and zero beta particles per alpha decay of 2"U, an alpha to beta ratio of 2.3 : I will be used. Table A-1 provides a summary of the I principle uranium and uranium progeny radiations. b I I l P I i I I I Initial Issue A-3 3/30/93

i l Figure A-1 Uranium Decay Series .I-for Radionuclides of Concern i i J I 9 l 235 231 p-231 l a Y (7.0E8y)" (25h) " I I l 238 234 234m 234 Pa (1.2 min) - U U (4.5E9y)- Th (24d) I 234 a 230 l U (2.4ssyy Th (ue4y) I I I I initial lssue A-4 3/30/93 I

Table A-1 Principle Uranium and Uranium Progeny Itadiations I Heta Particles per Alpha Activity Uranium Alpha Weighted #/a Radionuclide Fraction Decay Ratios l 23'U 0.77 235U 0.03 l 23'Th 1 0.03 238U 0.20 234Th 1 0.20 234*Pa 1 0.20 I Total 1.0 0.43 Total; a/# Ratio = 2.3 : 1 I I I I I I Initial Issue A-5 3/30/93 I

I I I Appendix B Ll Temporary Storage of Radioactive Waste I I 1 4 j i I I l j ) initial lssue B.1 3/30/93 j e )

Appendix H Temporarv Storace of Radioactive Waste Low-Level Radioactive Waste Low level radioactive wastes (LLRW) will be packaged in accordance with applicable regulations and delivered to a carrier for transport to an approved waste processor and/or disposal facility. LLRW packages awaiting shipment to a processor or disposal facility may be temporarily stored for up to one year on site. LLRW storage areas shall be appropriately posted and the LLRW shall be secured from unauthorized removal. Prior to being placed in storage, packages will be checked for exterior contamination and labeled as to enrichment and "U l 2 content. LLRW may also be stored for up to five years as interim storage provided that it is I-protected from the elements. Interim LLRW storage shall be in locked C-van trailers and in i shelters on site. Packages in interim LLRW sheltered storage will be checked annually for exterior contamination, and the adequacy of the package condition verified. The requirements for storage do not apply to waste which is in process in accordance with approved procedures. Records will be maintained of the contents of LLRW packages. All packages will be stored n on raised platforms (e.g., built in or portable pallets) and package stacking will be limited to three (3) high. When placed in storage, the packages shall be sealed in a manner which , f precludes casual entry (e.g., by the use of steel clips or strapping) until fmal sealing is accomplished prior to shipment to the processor or disposal facility. Packages containing liquid wastes shall be segregated from solid waste packages (e.g., stacked separately) and shall I be appropriately labelled. Liauid Wastes Clean-up rinse water solutions are sampled to verify that MPC is not exceeded, and are then released to the liquid waste system, which is controlled under a separate license. Release of lI liquid waste will be authorized by a member of the Radiological Protection and Industrial Safety staff. I I Initial Issue B-2 3/30/93 I

'i l 1 8 l l l i 1 I i h T l 4 k I t t I s Appendix C I i l Glossary l l i 1 I l I ? .i i l t 1 9 a 1 6 i 1 ] 4 t 6 I 1 i j i i Initial Issue C-1 3/30/93 s ll J,

i i t Appendix C Glowarv Activity: A measure of the rate at which radioactive material is undergoing radioactive I decay, usually given in terms of the number of nuclear disintegrations occurring in a given quantity of material over a unit of time. The unit of activity is the curie (Ci). Also, known as Radioactivity, I i Alpha Particle: A positively charged particle emitted by the nuclei of some radioactive elements. Alpha particles are identical to a helium nucleus. Annual Limit on intake (ALI): The derived limit for the amount of radioactive material I taken into the body of an adult worker by inhalation or ingestion in a year. ALI is the smaller value ofintake of a given radionuclide in a year by the reference man that would result in a committed effective dose equivalent of 5 rems or a committed dose equivalent of 50 rems to any given organ or tissue. Audit: Audits are examinations made to verify that operations are being performed according to established criteria. Background Radiation: Naturally occurring radiation in the human environment. It includes cosmic rays, radiation from the naturally radioactive elements, and man-made radiation from global fallout. Becquerel: A unit, in the international system of units, of the measure of radioactivity equal to one transfonnation per second. Beta Particle: A charged particle emitted from a nucleus during radioactive decay, with a mass equal to 1/1833 that of a photon. i Characterization Survey: Facility or site sampling, monitoring, and analysis activities to determine the extent and nature of contamination. Characterization provides the basis for acquiring the necessary technical information to develop, analyze, and select appropriate I cleanup techniques. I Cleanup: Actions taken to remove a hazardous substance that could affect humans and/or the environment. The term " cleanup" is sometimes used interchangeably with the terms Remedial Action, Remediation, and Decontamination. Committed Effective Dose Equivalent: The sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the committed dose equivalent to these organs or tissues. Initial issue C-2 3/30/93 I

Glossarv (cont'd) Confirmatory Survey: Limited independent (third-party) measurements, sampling, and I analyses to verify the findings of a final status survey. Contamination: The presence of residual radioactivity, in excess of levels which are I acceptable for release of a site or facility for unrestricted use. Criteria (release criteria): Combination of numerical activity guideline levels and conditions for their application. If criteria are satisfied, the site may be released without restrictions. I Curie: A measure of the rate of radioactive decay. One curie (Ci) is equal te 37 billion disintegration per second (3.7 x 10" dis's). which is approximately equal to the decay of one gram of radium-226. Fractions of a curie, e.g., picoeurie (pCi) or 10.i2 Ci and microcurie 4 (pCi) or 10 Ci, are levels typically encountered in the decommissioning process. Decay: The spontaneous radioactive transformation of one nuclide into a different nuclide or l l into a lower energy state of the same nuclide. Also, known as Radioactive Decay. Decommissioning: The process of safely removing a facility from operation and reduemg residual radioactivity to a level that permits release of the facility for unrestricted use. Decontamination: The removal of unwanted radioactive material from facilities, soils, or equipment. Also, known as Remediation, Remedial Action, and Cleanup. l Deep Dose Equivalent (DDE): DDE is the dose equivalent at a tissue depth of I cm I (1000mg/cm ). 2 i Derived Air Concentration (DAC): The concentration of a given radionuclide in air which I if breathed by the reference man for a working year of 2,000 hours under conditions oflight work (inhalation rate 1.2 cubic meters of air per hour), results in an intake of one ALI. Direct Measurement: Radioactivity measurement obtained by placing the detector against the surface or in the media being surveyed. The resulting radioactivity level is read out directly. Dose Equivalent (Dose): A term used to express the amount of effective radiation when I modifying factors have been considered. It is the product of absorbed dose (rads) multiplied by a quality factor and any other modifying factors. It is measured in ren. (roentgen equivalent man). l Elevated Area: Small, isolated location where radiation or radioactivity level is higher than the guideline level but satisfied other conditions. l I initial Issue C-3 3/30/93 i I I

I Glossarv (cont'd) .g Esposure Rate: The amount of ionization produced per unit time in air by X-rays or gamma g rays. The unit of exposure rate is roentgenstour (R/h); for decommissioning activities the 4 typical units are microroentgens per hour (pR/h), i.e.10 R/h. Final Status Survey: Measurements and sampling to describe the radiological conditions of a site, following completion of decontamination activities (if any) and in preparation for unrestricted release. Gamma Radiation: Penetrating high-energy, short-wavelength, electromagnetic radiation I (similar to X-rays) emitted during radioactive decay. Gamma rays are very penetrating and require dense materials (such as lead or uranium) for shielding. I. Grid: System of coordinates established on a site for purposes of referencing survey locations. Italf-Life: The time it takes for half the atoms of a quantity of a particular radioactive element to decay into another form. IIalf-lives of different isotopes vary from millionths of a second or less to billions of years. IIEPA Filter: liigh efficiency particulate filter. l Inspection: Inspections are routine reviews to ensure that operations are being conducted according to approved procedures. Investigation: Investigations are systematic examinations made to determine the cause(s) of, etTect(s) of, and/or corrective action (s) required due to abnormal occurrences. License: Authorization by NRC to possess, use, transfer, etc., radioactive materials for specified applications and under established conditions. Minimum Detectable Activity (MDA): The minimum level of radiation or radioactivity that can be measured by a specific instrument and technique. The MDA is usually established on the basis of assuring false positive and false negative rates ofless than 5%. Radionuclide: An unstable nuclide that undergoes radioactive decay. Release Criteria: Numerical guidelines for direct radiation levels and levels of radioactivity in soil on surfaces which are considered to be acceptable within a given set of conditions and applications. REM (Roentgen Equivalent Man): A quantity used in radiation protection to express the effective dose equivalent for all forms ofionizing radiation. It is the product of the absorbed initial Issue C-4 3/30/93

1 I-1 Glossarv (cont'd) dose in rads and factors related to relative biological effectiveness. Also, see Dose I Equivalent. Remediation: The removal of contamination from a site. Also, known as Itemedial Action and Decontamination. !g itemovable (Loose) Activity: Surface activity that can be removed and collected for !E measurement by wiping the surface with moderate pressure. Iloentgen (R): Unit of exposure. One roentgen is the amount of gamma rays or X-rays required to produce one electrostatic unit (esu) of charge of one sign (either positive or l negative) in one cubic centimeter of dry air under standard conditions. Scanning: An evaluation technique performed by moving a detection device over the surface at some consistent speed and distance above the surface to detect elevated levels of radiation. Scanning provides qualitative or semi-quantitative, rather than quantitative, data. Scoping Survey: A survey that is conducted to identify which radionuclides are present as j contaminants, relative ratios in which they occur, and the general levels and extent of the contaminants. ll Shallow Dose Equivalent: The external dose to the skin or extremity at a tissue depth of 0.007 cm. Soil Activity (Soil Concentration): The level of radioactivity present in soil and expressed in units of activity per soil mass [ typically picoeuries per gram (pCi/g)). Surface Activity: Radioactivity found on building or equipment surfaces and expressed in 2 units of activity per surface area [ typically disintegration per minute per 100cm 2 (dpm/100cm )]. Survey: Evaluation of a representative portion of a population to develop conclusions regarding the population as a whole. In the decommissioning process several different types of surveys are conducted, including llackground, Scoping, Characterization, Remediation Control, Final Status, and Confirmatory. Special Nuclear Material: Plutonium, U-233, and Uranium enriched in U-235. Special nuclear material is generally considered material capable of undergoing a fission reaction. Unrestricted Use: Use of a former radioactive materials site without requirements for future radiological controls. Also, known as Unrestricted Release. Initial Issue C-5 3/30/93 I I

I l l i i 9 l ] I i l 4 . l i Appendix D i ll References 1 d i l i i 1 I l 1 4 ll 1

I s

Initial Issue D-1 3/30/93 l l ^ 1 \\

Appendix D REFERENCES 1) NUREG / CR-2082 " Monitoring for Compliance With Decommissioning Termination Survey Criteria", Parts I and II. 2) NUREG / CR-5849 " Manual For Conducting Radiological Surveys in Support of License Termination" (Draft). 3) Branch Technical Position, " Disposal or Onsite Storage of Residual Thorium or Uranium (Either as Natural Ores or Without Daughters Present) From Past Operations" 4) SNM-1067 Annex B, " Guidelines for Decontamination of Facilities and Equipment I Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source or Special Nuclear Material". 5) Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors". 6) Regulatory Guide 4.15, " Quality Assurance for Radiological Monitodug Programs (Normal Operations) - Effluent Streams and the Environment". 7) Regulatory Guide 3.65, " Standard Format and Content of Decommissioning Plans for Licensees under 10 CFR parts 30,40, and 70" 8) Regulatory Guide 3.52, " Standard Format and Content for the Health and Safety Sections of License Renewal Applications for Uranium Processing and Fuel Fabrication" 9) Regulatory Guide 10.3, " Guide for the Preparation of Applications for Special Nuclear Material Licenses of Less Than Critical Mass Quantities". 10) "Developmcat and Use of Statistical Survey Criteria for Release of Materials at a Former Uranium Processing Facility", from Health Physics Society Journal, Volume 61,No.6. I1) Title 10, Code of Federal Regulations. 12) ANSI N13.1 - 1969, " Guide to Sampling Airbome Radioactive Materials in Nuclear Facilities" Initial Issue D-2 3/30/93 I}}