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33 Broad Street, 10th Floor, Boston, MA 02109 www.aeiconsultants.com Tel 857.350.3519 Fax 857.233.5531 Page 1

April 23, 2015 AEI Project No. 341384

Mr. Gerad Martin Mass Department of Environmental Protection Southeast Region20 Riverside DriveLakeville, MA 02347 [email protected]

Ms. Millie Garcia-Serrano, Deputy Regional DirectorMass Department of Environmental Protection Southeast Region 20 Riverside DriveLakeville, MA 02347

[email protected]

Subject: Transmittal of DRAFT Radiological Planning Documents BASF Facility, 32-42-Taunton Street, Plainville, MA 02762

Dear Gerad and Millie:

AEI Consultants was retained by BASF Corporation (BASF) to prepare the following DRAFT radiological planning documents, which are being submitted to you on their behalf, and are attached to this transmittal letter:

• Radiological Risk Assessment, Dated February 5, 2014;• Final Radiological Status Survey Plan, Dated April 4, 2014; and• Radiological Decommissioning Work Plan, Dated March 31, 2015.

BASF looks forward to your review and any comments of these documents. Please provide any questions and comments to Mr. Ed Vanyo at BASF, [email protected], (973)-245-5610.

Sincerely,

AEI Consultants

Stephen J. Graham, PE, LSP Senior Vice President

cc. E. Vanyo, Transmittal Only

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March 31, 2015

DECOMMISSIONING WORK PLAN

BASF Corporation32 Taunton Avenue, Plainville, MA

Prepared for:BASF Corporation100 Park AvenueFlorham Park, NJ 07932

Prepared by:AEI Consultants33 Broad Street, 10th FloorBoston, Massachusetts 02109(857) 233-5531

Decommissioning Work Plan, BASF Corporation, Plainville, MAMarch 31, 2015 Page 1

TABLE OF CONTENTS

1.0 BACKGROUND AND INTRODUCTION...................................................... 1 1.1 General .......................................................................................................... 1 1.2 Relevant Plant Operating History...................................................................... 1 1.3 Historical Radiological Investigations and Remedial Actions ............................... 2 1.4 Regulatory Overview ....................................................................................... 3

1.4.1 MADPH/RCP Restricted Conditions ................................................................. 3 1.4.2 MADEP Activity and Use Limitation ................................................................. 3

2.0 DECOMMISSIONING OBJECTIVES AND ACTIVITIES.............................. 4 2.1 Risk Assessment and Soil Cleanup Criteria ........................................................ 4 2.2 Planned Regulatory Decommissioning Activities ................................................ 4

2.2.1 Building 2 Slab ............................................................................................. 5 2.2.2 Building 2 Areas Room 2M and Tunnel ........................................................... 5 2.2.3 Building 12 Slab ........................................................................................... 6 2.2.4 Building 3 Slab ............................................................................................. 6 2.2.5 Upper Courtyard........................................................................................... 6 2.2.6 Drywell ........................................................................................................ 7

2.3 Planned ALARA Investigations and Remediation................................................ 7 2.3.1 Turnpike Lake .............................................................................................. 7 2.3.2 Dry Well Service Lines................................................................................... 7 2.3.3 Leach Field Piping......................................................................................... 7 2.3.4 Elevated Concentrations in Sub-Surface Soil.................................................... 8 2.3.5 Other Pending Investigations & Potential Remediation ..................................... 8

2.4 Description and Summary of AOC B, 7 & 16 Land Areas .................................... 8 2.4.1 Leach Field and Dry Well Land Area (LF/DW) .................................................. 9 2.4.2 Former Building 2 Sub-Slab Soils (B2SUB)....................................................... 9 2.4.3 Upper Courtyard and Area (UCY) ................................................................... 9 2.4.4 Lower Courtyard and Delta Sediment Land Area (LCY and SED) ....................... 9

2.5 Final Site Status Survey................................................................................... 9 2.6 Long Term Monitoring Plan & Institutional Controls.........................................10 3.0 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES ......... 11 3.1 General Requirements for Decommissioning Operations Contractor..................11 3.2 Preliminary Schedule of Decommissioning Activities ........................................12

3.2.1 Pre-Decommissioning Activities.....................................................................12 4.0 RADIATION HEALTH AND SAFETY ........................................................ 14 4.1 Plans and Procedures ....................................................................................14 4.2 Radiation Detection Instrumentation & Measurements ....................................15

4.2.1 Typical Portable Radiation Detection Instrumentation Used in Prior Investigations and Proposed Decommissioning...............................................15

4.2.2 Contamination Surveys for LEU.....................................................................15 4.2.3 Soil Sample Radioanalysis.............................................................................16

4.3 Personnel Training and PPE ...........................................................................17


4.4 Radioactive Waste Management & Disposal ....................................................17 4.5 Site Security ................................................................................................. 18 4.6 Quality Assurance .........................................................................................18 4.7 Final Site Status Survey and Report................................................................19 5.0 PUBLIC INVOLVEMENT IN THE DWP .................................................... 20 6.0 REFERENCES ......................................................................................... 21

Tables

Table 1 SDI Beta-Gamma Radiation Surveys UCY Soil Borings Table 2 ALARA Investigation LocationsTable 3 Summary of Remediation and Investigation AreasTable 4 Detector Conversion Factors for Total to Measured RadioactivityTable 5 Alpha Spectroscopy and Gamma Spectroscopy Comparison

Figures

Figure 1A AOC B, 7, and 16 Proposed 0-2’ Excavation ALARA InvestigationsFigure 1B AOC B, 7, and 16 Proposed 2-4’ Excavation ALARA InvestigationsFigure 2 Room 2M and Tunnel Soil RemovalFigure 3 Proposed Courtyard ExcavationFigure 4 Proposed Drywell ExcavationFigure 5 Proposed Sediment ExcavationFigure 6 Proposed Leachfield Investigations

Attachments

Attachment A Uranium Concentrations in Soils Building 2Attachment B Uranium Concentrations in Soils Upper CourtyardAttachment C Uranium Concentrations in Soils Lower Courtyard and SedimentAttachment D Uranium Concentrations in Soils Dry Well and Leach Field

Appendices

Appendix 1 Radiological Risk Assessment and Technical Basis for the Derived Concentration Guideline Value (DCGL)

Appendix 2 Site Final Status Survey Plan


DECOMMISSIONING WORK PLAN

1.0 BACKGROUND AND INTRODUCTION

1.1 General

This Decommissioning Work Plan (DWP) provides the scope of work, schedule, organization and quality assurance program for the decommissioning of radiologically-impacted soil, sediment, and other materials at the BASF Corporation site in Plainville, Massachusetts (the Site), and updates a prior DWP submitted to MADPH in 1997 (Engelhard, 1997). The exterior soils decommissioning program described herein represent the second and final phase of such effort at the Site. A Nuclear Regulatory Commission (NRC) - approved plan (Engelhard, 1996) had already been conducted through March 1997 to decontaminate and decommission Building 2, and upon NRC approval (NRC 1997), this building was dismantled to the floor slab.

The Site is also undergoing a separate environmental cleanup program under the authority of the U.S. Environmental Protection Agency (EPA). This separate program is being implemented under the federal Resource Conservation and Recovery Act (RCRA) Corrective Action Program, and involves remediation of soils and sediment with elevated non-radiological constituents of concern which, in some instances, are physically co-mingled with radiological contamination. Remediation for non-radiological contamination is not addressed in this Plan.

1.2 Relevant Plant Operating History

Engelhard Corporation (previously the Makepeace Division of Engelhard Industries, Inc.) manufactured nuclear fuel elements at the Site from 1957 until cessation of operations in 1962. This manufacturing plant operated under a license with the federal Atomic Energy Commission (AEC), the predecessor agency to the NRC. Manufacturing operations principally involved the use of low-enriched uranium in Building 2, where fuel element fabrication was conducted. Fuel element fabrication operations were completely separate from non-nuclear manufacturing and other facilities (Engelhard 1995).

Existing documentation is limited concerning the specific practices and procedures used in handling solid wastes generated from the nuclear operations. However, information obtained from interviews with the Plant Engineer and Health Physicist from that period revealed that all radiologically contaminated solid waste produced in the nuclear operations were collected and placed into 55 gallon drums. These filled solid waste drums were then temporarily stored in a fenced storage area in the "courtyard" area west of Building 2 until off-site disposal was undertaken. No waste was disposed on-site (Engelhard 1995).

An AEC-approved incinerator was constructed and operated on the site starting approximately in March 1960. A subsurface leach field disposal system to treat nuclear process waste water had also been installed at the plant in 1957. The leach field system was modified over time, and eventually consisted of two pits which distributed water to the leach field. The system was designed to handle treated, monitored and decontaminated wastes from uranium processing, and from laundry, shower, and laboratory wastes. The leach field system was monitored to verify that radionuclide concentrations in discharges to the leach field were within former AEC release standards in 10 Code of Federal Regulations (CFR) Part 20 (Engelhard 1995).


No drawings showing the as-built location of the components of this system have been found. However, it is known that in 1960, Engelhard obtained approval from the Massachusetts Department of Public Health for expansion of the liquid waste disposal system by the addition of a second leaching pit. In August 1961, it was found that a blocking of one of the pits by a scum from the liquid waste caused waste liquid to overflow. This problem was apparently corrected by the addition of sulfuric acid, which cleared the blockage. Further corrective action, consisting of the installation of three parallel "leaching trenches," running from west to east was planned to prevent future overflows from the site. There is no documentation available to confirm whether the trenches were constructed, or to what length if constructed, but the existence of a distribution system in the region of the site has been presumed. A high water alarm was also installed in the site. There is no indication of any further problem with the system (Engelhard 1995).

Use of the system for nuclear waste disposal apparently ended with the cessation of nuclear operations in 1962, although the same system may have continued to be used for non-nuclear waste water disposal until 1972, when industrial waste water was diverted to a new treatment system, or until 1976 when shower and sink wastes were diverted to a sanitary sewer(Engelhard 1995).

A dry well, lined with cinder concrete blocks and approximately eight feet deep, was installed at some point during the history of Engelhard operations. It is not known if it was operational during the time of the nuclear activities but it is possible that a piping connection between those operations and the dry well existed then or perhaps later. This connection could have been an avenue for discharge of water exhibiting elevated radioactivity to enter the dry well(Engelhard 1995).

A Final Radiological Survey was conducted by Engelhard in 1963. This Survey was undertaken subsequent to decontamination operations, and served as the basis for requesting a license termination, prior to a AEC field validation survey. The Survey covered the Building 2 interior and equipment. No data was available from that time on radioactivity in environmental media external to Building 2 or its exterior facility surfaces. The facility was granted a license termination by the AEC in late 1963 (Engelhard 1995).

1.3 Historical Radiological Investigations and Remedial Actions

The following programs have been undertaken to date at the Site:

• A radiation scoping survey of the Site was performed in July 1988 as part of a multi-phase site assessment initiated by Engelhard to identify area of environmentalconcern. The survey identified radiological contaminants of concern in exterior andinterior of Building 2 at the Site (Engelhard 1995).

• Detailed characterization surveys of Buildings 1 & 2 were conducted in March 1994which resulted in the delineation of impacted areas based on the NRC regulatoryguidelines. An NRC-approved Decontamination Plan (Engelhard, 1996) forunrestricted release of former Buildings 1 & 2 was implemented in 1996. The NRCapproved the Final Status Survey of Building 1 & 2 in 1997 (NRC, 1997). Thedecommissioned buildings 1 & 2 were later dismantled. Soils under the formerBuilding 2 slab footprint were remediated in this project with the exception of asmall volume of soil subject to further assessment with the remaining exterior


contaminated soils at the Site. • A comprehensive Site Characterization Program (SCP) of exterior areas of the Site

had been completed earlier in 1995 (Engelhard, 1995). The SCP generated detailed information on the distribution of radiologically contaminated soils exterior to former Building 2.

• Additional radiological data was obtained during performance of a Preliminary DesignReport (Engelhard, 2001) and Supplemental Investigation in the Courtyard area of Areas of Concern (AOC) B/7 in 2004 (Engelhard, 2004).

• Packaging, transport, and off-Site disposal of former Building 2 decommissioningwaste materials from the Building 11 Radioactive Materials Storage Area and Final Status Survey of Building 11 was accomplished in 2010 (Environ, 2010).

1.4 Regulatory Overview

MADPH/RCP Restricted Conditions1.4.1The Massachusetts Department of Public Health Radiological Control Program (RCP) became the regulatory authority for radiological concerns upon achieving NRC Agreement State status in 1997. BASF intends to demonstrate in this DWP how it will both meet RCP regulations, and achieve the RCP restricted criteria (only future industrial or commercial use) of 105 CMR 120.246 ‘Criteria for License Termination Under Restricted Conditions’ while allowing certain residual radioactive soils to be left on-site. BASF will make provisions for legally enforceable institutional control (environmental deed restriction) that provides assurance in accordance with well-accepted Massachusetts protocols, that the total effective dose equivalent (TEDE) concentration remaining from residual radioactivity (that is distinguishable from background) from the average member of the maximally exposed receptor (Site worker), will not exceed the RCP regulatory limit of 10 mrem per year.

MADEP Activity and Use Limitation1.4.2The Site is also currently regulated by the Massachusetts Department of Environmental Protection (MADEP) under the Massachusetts Contingency Plan in accordance with 310 Code of Massachusetts Regulations 40.00 as a site which has released “oil and hazardous materials, or OHM”. BASF intends to utilize a MADEP-administered environmental property deed restrictiontool, known as an Activity and Use Limitation (AUL), which a legal document that restricts future Site uses to commercial or industrial. The AUL specifies site conditions and property management practices that must be maintained to allow a condition of No Significant Risk (NSR) to be achieved. The NSR condition allows concentrations of OHM remain at levels above regulatory standards for current or future use as a residential or sensitive receptor use, yet will exist at or below regulatory standards for current or future industrial or commercial use of a property. The AUL itself does not create a condition of No Significant Risk; it is used to minimize the chance of an unforeseen change in the Site use that could result in unacceptable exposure to chemical contaminants. The AUL instrument is accepted by the RCP and EPA for the purpose of legally demonstrating that future site use will be restricted to non-intrusive activities, and to prevent unacceptable exposures to relatively low levels of contaminants that may be left at the site.

Decommissioning Work Plan, BASF Corporation, Plainville, MAMarch 31, 2015Page 4

2.0 DECOMMISSIONING OBJECTIVES AND ACTIVITIES

2.1 Risk Assessment and Soil Cleanup Criteria

The RESRAD computer code, Version 6.3, 2005, has been used to perform pathway analyses for land areas at the Engelhard Site that have been impacted by the former Building 2 nuclear facility operations. RESRAD is a computer code developed by Argonne National Laboratory (ANL), and recognized by the NRC and MADPH, to calculate site-specific residual radioactive material guidelines as well as radiation dose and excess lifetime cancer risk to a theoretical, critically exposed on-site occupant. The technical basis for determination of the Site’s residual radioactive material guideline has been previously prepared in a document entitled Radiological Risk Assessment and Technical Basis for the Derived Concentration Guideline Value or DCGL (AEI, 2014a) and presented as Appendix A to this DWP.

Dose limits are typically stated as total effective dose equivalent (TEDE), while risk is usually stated in terms of morbidity (probability of cancer incidence) or mortality (probability of cancer death). Whether stated in terms of dose or risk, such release limits generally cannot be measured directly. As a consequence, pathway analyses are typically performed to translate dose and/or risk into equivalent environmental concentrations. These environmental concentrations are called Derived Concentration Guideline Levels (DCGLs). The DCGLs have been analytically determined by evaluating a Site industrial/commercial future use scenario wherein the exposure pathways are direct radiation exposure and ingestion/inhalation of radioactive materials.

The DCGL values are radionuclide specific; the numerical value of each area-specific radionuclide concentration is defined in units known as picoCuries/gram (pCi/g) homogeneously distributed in the contaminated zone, and is equivalent to a TEDE of 10 mrem/yr for the above-referenced exposure pathways calculations. Multiple radionuclide guideline values are desirable when mixtures of radioactive material contribute to changes in the relative levels of principal radionuclides, i.e., when the ratios of the principal uranium isotopes are variable, as is the case at this Site.

Sample results historically collected on Site and analyzed for the principal uranium isotopes need to be compared to the DCGL release criteria by the Sum Fraction Unity Rule (AEI, 2014). The Unity Rule is satisfied (is less than the release criterion TEDE) when radionuclide mixtures yield a combined fractional concentration limit that is less than or equal to one.

The following isotope specific DCGL values for U238, U235, and U234 have been determined as the Site soil guideline values based on RESRAD composite site exposure pathway calculations:DCGLU238 = 98 pCi/g; DCGL U235 = 22 pCi/g; and DCGL U234 = 595 pCi/g (AEI 2014a).

2.2 Planned Regulatory Decommissioning Activities

Areas of Concern (AOC) B, 7, 16, and the former Building 2 floor slab, also called the radiologically impacted area, are shown on Figures 1A and 1B. These areas require either remediation or a demonstration that RCP regulatory levels have been met. Each of these areas is considered in more detail below.


Building 2 Slab2.2.1The former Building 2 floor slab and adjacent Tunnel represents the limits of the decommissioning and unrestricted release area for the former uranium fabrication facility, asapproved by the NRC (NRC, 1997), with the exception of limited sub slab soils. A separate designation was made for small pockets of soil underlying the Tunnel, Room 2 M and the adjacent foundation wall at the completion of the Building 2 decommissioning. These localized ‘special soil areas’ contained concentrations of uranium greater than the 1996 NRC guideline criteria and were (and remain) located at a depth greater than 4’ below the existing Building 2 slab. Because of the depth and proximity to the exterior, Engelhard with NRC concurrence, grouped these areas with the exterior soils for later disposition. These soil areas exceed the current soil cleanup criteria and are planned for remediation (see Section 2.2.2).

The Building 2 slab is a high density 8” concrete thickness of approximately 2800m2 and the slab elevation is about 4’ above existing grade. The currently exposed slab surface was decontaminated and approved for release during the 1996 decommissioning. All of the sub slab drain lines and contaminated expansion joints were removed during the decommissioning (Figure 2, Attachment A), and the floor cuts are still visible on the slab. The slab cuts were backfilled with clean material and debris after piping removal, soil remediation, and confirmatory sampling. The release criteria for soil in 1996was 30 pCi/g total Uranium radioactivity. Therefore, it is not expected that any soils within these slab cuts will exceed the current DCGL criteria. Remaining soils under the intact floor slab are not expected to contain radioactive materials from former uranium fabrication operations.

Floor slab concrete and foundation materials overlying radiologically contaminated soil and now planned for removal, are not expected to be contaminated. Such materials will be evaluated via radiological instrumentation (henceforward termed “evaluated” in this DWP), remediated (if required), and excavated and transferred to a temporary holding area external to AOC B, 7 and 16, pending completion of radiological activities.

All soils under the former Building 2 slab which is planned for removal to grade, will be evaluated for compliance with the soil criteria. Based on the need to evaluate an approximate 3-4’ soil depth under the slab, about 3500 cy of soil will be evaluated (see delineated area in Figure 2), remediated (if required), and excavated and transferred to a temporary holding area external to AOC B, 7 and 16 pending completion of radiological activities.

Building 2 Areas Room 2M and Tunnel2.2.2Three sub-slab soil areas were left at the conclusion of the Building 2 decommissioning for later remediation: the excavation in Room 2M (southwest corner of former Building 2) and two areas in the west end of the Tunnel (Figure 2). The Tunnel was a maintenance access corridor to the waste processing system located at the east end; the floor surface of the Tunnel sloped from ground level at the west end entry to >8’ bgs at the east end, where a wastewater treatmentprocess operation was located. Significant excavations were performed in these areas, particularly at the west end (Room 2M) and west Tunnel. Confirmatory soil sampling by the NRC and Engelhard’s radiological contractor in 1996 confirmed the presence of soil contamination that exceeded the former soil cleanup criteria. These samples have been evaluated for compliance to the current soil cleanup criteria and determined to be greater than allowable concentrations. A 15 square foot of contaminated concrete greater than the 5000 dpm alpha per 100 cm2 guideline value for surfaces also exists in the tunnel. This surface is at


the interface between the floor and a concrete pad at the far eastern end of the Tunnel. This surface lies at as depth greater than four feet below the Building 2 slab, but may be encountered during soil excavation.

These three soil area locations (Figure 2) are to be remediated to a depth of 4’ bgs (below Site grade) either at the time of Building 2 slab removal or subsequent to slab removal dependent on the access to these areas.

Attachment B presents the results of soil samples acquired during the Building 2 decommissioning. Soil samples marked as >DCGL in the DCGL test column exceed the current soil cleanup criteria.

Building 12 Slab2.2.3Building 12 was constructed after former Building 2 nuclear fabrication activities and had been determined to contain radioactive materials (Engelhard, 1995) probably due to backfill of the foundation from contaminated Site soils (Figures 2, 3). The Building 12 slab has a 6-12” concrete thickness of approximately 600m2 and the slab elevation is about 4’ above existing grade. This slab is unusually thick due to its former need to support a two-story ball mill.

All sub slab soils under former Building 12 planned for removal to grade will be evaluated for compliance with the soil criteria. Based on an approximate 3-4’ soil depth under the slab, about 700 cy of soil will be evaluated, remediated (if required), and excavated and transferred to a temporary holding area external to AOC B, 7 and 16 pending completion of radiological activities.

Building 3 Slab2.2.4Building 3 was constructed after former Building 2 nuclear fabrication activities and had been determined to contain radioactive materials (Engelhard, 1995) probably due to backfill of the foundation from contaminated Site soils (Figures 2,3). The Building 3 slab is a 6” concrete thickness of approximately 250m2 and the slab elevation is about 4’ above existing grade.

All sub slab soils under former Building 3 planned for removal to grade will be evaluated for compliance with the soil criteria. Based on an approximate 3-4’ soil depth under the slab, about 300 cy of soil will be evaluated, remediated (if required), and excavated and transferred to a temporary holding area external to AOC B, 7 and 16, pending completion of radiological activities.

Upper Courtyard2.2.5Uranium contamination in soils in the ”upper courtyard” (northern part of AOC B/7, Figure 3) had initially been determined during the Site Characterization (Engelhard, 1995). A later investigation (Engelhard, 2004) was performed to provide additional data on the presence of constituents of concern, including VOC's, PCB's, and metals, in the vicinity of the Upper Courtyard. The Courtyard is a known Site location where radioactive material is present andmixed with RCRA constituents of concern.

Soil contamination was detected by beta-gamma soil surveys in many of the borings and test pits. Two samples acquired for isotopic uranium analysis indicated soil concentrations that exceed the current soil cleanup criteria. The area of potentially impacted area of radiological


contamination in the upper courtyard is approximately 50’ x 160’ or 8000ft2 and extends generally from 1’ to 3’ bgs. A summary of the survey data is provided in Tables 1 and 2.

Figure 3 is a presentation of the sampling locations and site grid overlay and indicates the relative location of the test pit excavations performed during the second round of investigation. Attachment B presents the summary of soil samples acquired during the SCP and SDI for the Upper Courtyard. Soil samples marked as >DCGL in the DCGL test column exceed the current soil cleanup criteria.

Drywell2.2.6A drywell characterized during the SCP exists near the southeast corner of former Building 2, and was defined as AOC 16 under the separate RCRA investigation (Figure 4). MADEP regulations require that inactive dry well structures be excavated and otherwise decommissioned as part of remedial activities. This dry well structure is additionally considered radiologically contaminated and is planned for removal as part of the radiological remediation; one soil sample near surface exceeds the DCGL criteria (Attachments C, D). The dry well structure will be managed as radioactive material for removal operations and disposal. The excavation will be evaluated for compliance with the soil cleanup criteria to a depth of 4’ bgs.

2.3 Planned ALARA Investigations and Remediation

Additional remediation activities in addition to the regulatory requirements will also be implemented during site closure activities to be responsive to the NRC and RCP regulatory philosophy known as ALARA (As Low As Reasonably Achievable) cleanup, as described further below.

Turnpike Lake 2.3.1A small ‘delta’ of uranium contaminated shoreline sediment was found during the SCP; this was probably due to surface run-off from the Site soils via a storm water drain (Figure 5). BASF intends to install sheet piling barriers and remove approximately 30 cy of low level contaminated sediment from this area.

Attachment D presents the summary of sediment samples acquired during the SCP. Soil samples marked as >DCGL in the DCGL test column exceed the current soil cleanup criteria.

Dry Well Service Lines2.3.2BASF will remove the contaminated drywell located at the southeast corner of former Building 2, and investigate potential service piping from the southeast corner of the Tunnel to the drywell within 0-4’ bgs and remove if found (Figure 2).

Leach Field Piping 2.3.3BASF will investigate and remove the leach field areas where impacted (see Section 2.4) and under ALARA the possible presence of service piping, if any, will be evaluated within 0-4’ bgs in the assumed discharge origination point at the east wall of the former Building 2 tunnel (Figure 2).


Elevated Concentrations in Sub-Surface Soil2.3.4Several areas, primarily in the lower courtyard (Figure 2), were originally determined in the SCP (Engelhard, 1995) to have elevated radioactive material in the 0-4’ bgs soils. However, these areas have been re-evaluated during preparation of this DWP (AEI, 2014a, Appendix 1)and found not to exceed the soil cleanup criteria. Nonetheless, some or all of these areas may need to be remediated based on review and adequacy of the prior soil sampling distribution in these areas. Evaluation and/or remediation of one or more of these elevated areas are proposed to increase confidence for performance of the post-remedial action Final Status Survey Plan (Appendix 2), which is required for compliance with the Multi-Agency Radiation Survey and Site Investigation Manual, or MARSSIM (NRC, 2000), and therefore a RCP-required closure document, which provides final verification that RCP regulatory levels have been achieved. Investigation areas for the radiologically impacted soils are shown on Table 2.

Other Pending Investigations & Potential Remediation2.3.5A major investigation is currently planned under the RCRA program for AOC B, 7 & 16 to complete characterization on the nature and extent of impacted soils from VOCs (in groundwater) and metals (soil and groundwater) and PCBs (soil) remaining on-Site from historical operations. Over 100 soil borings are proposed to assess RCRA contaminants in the Radiologically Affected Areas; the soil boring plan encompasses areas previously assessed for radiological contaminants in the SCP (Engelhard, 1995), and will also provide additional radiological characterization data to supplement data gaps for subsurface soils.

Additionally, a final Gamma Radiation Site Surface Survey (GRSSS) is scheduled to be performed prior to Site radiological remediation activities, which will supplement an initial GRSSS which was completed in Fall 2014, and is still in preparation. A gamma surface survey provides additional information on the radiological condition of near surface soils where no sampling data is available or at the known areas of elevated subsurface contamination.

Results of these pending investigations are to be incorporated into the DWP and associated project documents prior to implementation of Site radiological closure activities.

Table 3 provides a summary of both regulatory remediation areas and investigation areas in the radiologically impacted areas.

2.4 Description and Summary of AOC B, 7 & 16 Land Areas

The environmentally impacted land areas of the Site have been previously identified and well characterized for radiological constituents in surface and subsurface soils (Engelhard, 1995). This land area is not intended for remediation, unless otherwise identified below, and has been segmented as follows for subsequent discussions below on radiological data distribution:

• Leach Field and Dry Well • Former Building 2 Sub-Slab • Upper Courtyard• Lower Courtyard and Delta Sediment (Turnpike Lake)

The location of the radiological Areas of Concern are described further below and shown on Figures 1A, 1B.


Leach Field and Dry Well Land Area (LF/DW)2.4.1This is the area east of the former Building 2 enclosed by the security fence on the north, east, and south and west boundary on a line assigned extending north-south of the former Building 2 east side. This area is characterized by the historical presence of the Building 2 processed effluent discharge leach field, former leach pits, potential discharge piping, and a dry well approximately 8 feet deep. A discussion of the operating history of the leach field was provided above. Of this area, only the inactive dry well structure is planned for excavation and removal in accordance with Massachusetts Dry Well/UIC Closure requirements during the site remediation, although other areas will be evaluated and remediated as needed, as described in Section 2.3and Figure 6.

Former Building 2 Sub-Slab Soils (B2SUB)2.4.2This is the soil under the existing concrete slab of former Building 2. During the Building 2 decommissioning, cuts were made in the floor slab to access and remove sub-slab drain systems and perform final surveys. Contaminated soil greater than 30 pCi/g total uranium (release criteria in use at the time of Building 2 decommissioning) was excavated and the slab cuts back-filled with clean rubble/soil. The slab footprint area also encompasses the former area known as the Tunnel at the south end of former Building 2. Packaged radioactive waste materials from the former Building 2 decommissioning were stored in a Radioactive Materials Storage Area (Building 11) and later transported to a licensed and RCP approved disposal facility (Environ, 2010). The areas to be evaluated and remediated as needed, are described in Section 2.3.

Upper Courtyard and Area (UCY)2.4.3This is an area bounded on the west by the former Building 6 slab, on the north by the former Building 9 slab, and on the east by the former Building 2 slab. The south boundary is on an assumed line extending east-west on the south end of former Buildings 6 and 2. Former Building 12 and Building 3 floor slabs are within this area. The upper Courtyard is characterized by a surface that is approximately 4 feet below the existing adjacent building floor slabs. Site radiological contaminants were found in this courtyard soil area during a supplementary Site investigation (Engelhard, 2004) and probable radiological contaminants exist within the fill soils under former Building 12 and Building 3 slabs, as noted in Section 2.3.

Lower Courtyard and Delta Sediment Land Area (LCY and SED)2.4.4This is a continuation of the upper Courtyard surface to the south end of the Site at Turnpike Lake. There is also a small sediment region at the margin of Turnpike Lake included in this area where Site soil contamination was deposited due to erosion and runoff. The west boundary extends just beyond the west edge of the former Building 6 slab and the east boundary terminates at the Leach Field Dry Well area. Only the 30 CY sediment area described in Section 2.3.1 will need to be remediated at this time.

2.5 Final Site Status Survey

A Final Status Survey as was performed for Building 2 decommissioning approval (NRC, 1997) is typically performed to demonstrate that residual radioactivity satisfies the predetermined soil cleanup criteria. Site AOC areas B, 7 & 16 represent a MARSSIM Class 1 land survey area in which soil contamination is known or expected. The Final Status Survey Report to be prepared for this Site closure (Appendix 2) will identify the survey unit designations, radiological


sampling/survey requirements and statistical tests required to evaluate data from final field status surveys and verify regulatory levels have been achieved. Site contaminants (Uranium) are present in natural background and statistical tests will use the Wilcoxon Rank Sum (WRS) statistical test. A final Site closure residual radiation, or ResRad, dose calculation will be performed with data input from the final status field survey to document the TEDE after post-closure, that is, after soil removal has occurred. The technical approach for the postremediation final status survey is presented as a supporting document to the DWP and shown as Appendix 2.

2.6 Long Term Monitoring Plan & Institutional Controls

The RCP regulations require the Site owner to provide sufficient assurance to enable a responsible government entity or independent third party, including a government custodian of a site, both to carry out periodic rechecks of the site no less frequently than every three years to assure that the institutional controls remain in place as necessary to meet the criteria of 105 Code of Massachusetts Regulations (CMR) 120.246(B) and to assume and carry out responsibilities for any necessary control and maintenance of those controls.

Property restrictions and maintenance for the Site will be developed by BASF in concert with the MADEP Activity and Use Limitation (Section 1.4.2) for oil and hazardous materials and will encompass the following activities:

• Restrictions for excavation and transfer of soils; future construction or modifications in this area will require radiological monitoring.

• Perform periodic radiological measurements at reproducible locations.• Continue the groundwater monitoring program.• Acquire periodic sediment samples from Site effluent streams for isotopic uranium

analysis. • Routine maintenance of the site and maintain integrity of the existing 8 foot high

galvanized steel fence surrounding the Site.


3.0 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES

3.1 General Requirements for Decommissioning Operations Contractor

This decommissioning program will be conducted by BASF with assistance of a selected team of environmental and engineering staff, functioning as the on-site Program Manager (PM) and Radiological Engineer (RE). The PM will supervise the activities of the decommissioning contractor, audit their health and safety operations, and verify that the proposed work scope is achieved. The PM and BASF will be the point of contact with the regulatory agencies. The BASF team will provide on-site technical and QA/QC support, assist in the selection of contractors, perform surveys, interpret results, and prepare the Final Status Report.

A decommissioning operators contractor (DOC) will be formally selected who is experienced in performing comparable decommissioning tasks under State and/or NRC guidelines and whose personnel have the appropriate radiological and OSHA health and safety training. The selected contractor will be required, at minimum to:

• Commit qualified and experienced key personnel to the project for its duration.• Provide a detailed field operations plan and schedule.• Demonstrate how the criteria in the DWP will be achieved and verified.• Demonstrate how EPA criteria for the management of non-radiological contaminants

will be met.• Provide a RCP and OSHA-compatible health and safety plan adapted to the project

conditions.• Establish site-specific QA/QC procedures and guidelines.

The contractor hired to perform the decommissioning activities will be required to have the minimum key personnel in the contractor organization is as follows:

Project Manager. 15 years or more experience in radiological and hazardous waste decommissioning projects including excavation and site restoration activities and familiarity with radiological decommissioning regulations, guidelines, and procedures. Bachelor's degree in engineering, science, or construction management (or equivalent training and experience).

Superintendent. 10 years or more experience in directing laborers and operators in excavation, and site earth moving, restoration projects. Familiarity with radiological, OSHA, and RCRA decommissioning programs and procedures.

Radiation Protection Supervisor. 10 years or more experience in radiological decommissioning projects. Thorough knowledge and experience in radiological health and safety programs. The RPS will direct the radiological controls personnel and be responsible for implementation of the contractor’s radiological controls program. Bachelor's degree in nuclear engineering or radiation health physics (or equivalent training and experience).

Health & Safety Supervisor. 10 years or more experience in RCRA and radiological decommissioning projects. Maintain the contractor’s radiological, occupational, and chemical safety programs. Bachelor's degree in discipline (or equivalent training and experience).


Each onsite worker will have OSHA 40-hour HAZWOPER training. Additionally, workers will receive site specific-safety training consistent with RCP requirements. This training will include basic radiation safety and site-specific procedures for radiation and contamination controls. Radiological controls personnel shall have appropriate training and experience for decommissioning projects.

All contractor personnel with the potential for exposure to radiation and contamination in the work areas will participate in a dosimetry program. This program will include thermo-luminescent detector (TLD) badges for external exposure potential and baseline/final laboratory bioassay for internal dose assessment.

The Site decommissioning will be conducted and monitored to assure the health and safety of the decontamination workers, other on-site personnel, and the public. Procedures will be followed to minimize the risk of injury, and to verify that exposure to any potential radiation exposure is controlled to ALARA levels. The concept recognizes that there are practical limits to the extent of reduction in radiation exposure which can and needs to be achieved based upon the exposure potential and the ability of technologies to further reduce such exposure.

This concept will be employed at both the level of the Program Manager and the decommissioning contractor. The BASF site management team will provide management oversight in health and safety. This will involve joint development with the contractor and approval of the health and safety plan, establishing and maintaining access control to the work areas, and having audits performed of ongoing health and safety programs. The audits will verify that contamination control and radiation control support surveys are conducted properly and that the protective work rules established for the project are adhered to in order to meet ALARA levels.

3.2 Preliminary Schedule of Decommissioning Activities

The general radiological sequence of work is expected to be as follows:

Pre-Decommissioning Activities 3.2.1BASF:

• Conduct a series of interactive workshops focused on radiological and RCRA aspects of Site remediation in coordination with MADPH. BASF intends these forums to provide an informal exchange of information with the community, prior to initiation of the formal MADPH public participation program identified in 105 Code of Massachusetts Regulations (CMR) 120.246(D)(1) and (2), and applicable USEPA RCRA requirements for public participation. BASF will use a combination of its in-house communication resources and local external public participation expert(s) to determine stakeholder issues, concerns, and methods of public education, such as fact sheets, mailings, and workshops.

• Clear AOC B, 7, & 16 of foliage, brush, and miscellaneous debris to allow unimpeded access to all areas destined for survey activities.

• Re-establish the Site survey grid (10m x 10m) and Survey Unit designations, inaccordance with the Final Status Survey Plan, using GPS/GIS technology.

• Site gamma radiation Site Surface Survey using GPS/GIS technology.• Update BASF internal planning document, the Radiological Field Operations Plan, or


RadFOP (AEI 2014b) and DWP based on radiological data obtained from the gamma radiation Site Surface Survey and other Site investigations, if any, performed prior to final Site radiological decommissioning.

• Prepare final Scope of Work and associated documents for RFP for issue to contractors.

• Select Contractor.

Decommissioning Contractor:

• Remove Building 2, Building 12, and Building 3 floor slabs and foundations to Site grade. Remediate soils in Building 2 (Room 2M and Tunnel) that exceed current soil cleanup criteria. Evaluate, segregate as required, and excavate soils from within foundations.

• Sequentially excavate and remediate soils from upper courtyard, drywell, and Turnpike Lake sediments. Evaluate all removed soils for compliance and segregate as required.

• Investigate and remediate subsurface process lines as directed by statement of work (SOW) prepared by BASF. Evaluate all removed soils for compliance and segregate as required.

• Excavate and remediate other elevated Site soils as directed by SOW. Evaluate all removed soils for compliance and segregate as required.

• Perform continuous remediation status surveys and final confirmation surveys/sampling for above remediation activities. Package or otherwise control contaminated soils destined for off-site disposal. Complete sampling for RCRA and radiological waste characterization (note: disposal facilities acceptance criteria are based on a priority of radiological characterization first, then other constituents of concern e.g., PCBs, VOCs, metals are taken into account).

• Return compliance soils removed and evaluated from building foundations to origin. • Perform Site final status survey and report. Final ResRad calculation based on post-

decommissioning field conditions.


4.0 RADIATION HEALTH AND SAFETY

4.1 Plans and Procedures

A comprehensive radiological and occupational health and safety program will be developed for the decommissioning activity. The program will encompass worker protection from construction hazards, chemical contaminants, and radiation and radioactive material. Standard Operating Procedures (SOPs) for field decommissioning operations shall be included when required. The components of the health and safety program will include the following:

• Definition of contractor organization and responsibilities and communication protocols with Owner/Owner representatives.

• Activity hazards analysis for radiological, physical, and chemical hazards.• Worker training inclusive of general health and safety training, OSHA training for

hazardous waste, site specific radiological and hazards training, and plan of the day (POD) safety and operations training.

• Access controls to the radiologically controlled area, including delineation of entry and egress methods.

• Site safe work practices.• Management commitment and practices to maintain ALARA program.• Personnel monitoring for radiation and contamination.• Occupational and environmental (property boundary) airborne particulate

monitoring.• Radiation and contamination controls; dust mitigation practices for excavation

activities. • Waste segregation and minimization strategy. • Emergency Plan. First Aid and/or medical treatment for accidents and injuries.

Material releases and operations shutdown criteria.• Record keeping, logs and reports, forms and documentation, responsibilities.


4.2 Radiation Detection Instrumentation & Measurements

Typical Portable Radiation Detection Instrumentation Used in Prior 4.2.1Investigations and Proposed Decommissioning

Measurement Instrument DetectorBeta-gamma integrated surface survey

Ludlum M2221 Ludlum M44-9 GM pancake detector (15.5 cm2)

Beta-gamma surface survey

Ludlum M2/M3 Ludlum M44-9 GM pancake detector (15.5 cm2)

Gamma radiation surveys (uR/hr)

Ludlum M19 NaI(Tl) scintillation detector (internal)

Beta-gamma/alpha integrated smear measurements

Ludlum M2929 ZnS / plastic scintillator

Gamma radiation surveys (cpm)

Ludlum M2221 Ludlum M44-10 gamma NaI (Tl) scintillation

Beta-gamma surface survey

Ludlum M2221 Large Area Gas Flow Proportional Detector (425 cm2)

Beta-gamma personnel surveys

Ludlum M177 Ludlum M44-9 GM pancake detector (15.5 cm2)

Contamination Surveys for LEU4.2.2The radioactivity of processed uranium involves the principal uranium isotopes of U234, U235, and U238, and the immediate short lived progeny of these isotopes, i.e., U238 progeny Th234 and Pa234m, and U235 progeny Th231. These progeny of U238 and U235 are in secular equilibrium and can be used as a measure of the specific uranium isotope concentration. A beta particle measurement conversion to total uranium (i.e. total alpha) in disintegrations per minute, or dpm, depends upon uranium isotopic composition (since no progeny for U234 contributes to the total beta emission). Since each of the uranium isotopes emits one alpha particle per disintegration, an alpha radioactivity measurement is independent of isotopic composition.

The effectiveness of an alpha survey is very dependent on alpha self-absorption, surface conditions, roughness, contaminant distribution in surface layers, etc.; therefore it is not the best choice of methodology to ensure assessment or remedial objectives are satisfied. Beta-gamma measurements can provide a greater degree of confidence that objectives have been met. However, a detector calibration conversion factor must be used to account for the contribution of all uranium isotopes, i.e., total alpha radioactivity.

Radioactive contaminants at the Site have been identified in earlier investigations as processed uranium used in the fabrication of fuel elements during the period of Building 2 nuclear operations. The processed uranium is both the normal isotopic uranium composition and low level enriched uranium of approximately 4% U235 percent by weight (wt%). Conversion factors have been developed (Engelhard, 1996) to convert the measured (β-γ) radioactivity to total uranium (α) radioactivity for comparison to the NRC regulatory guidelines shown below.


Uranium Surface Contamination Guideline1 Values

Measurement Type Regulatory Criteria Average acceptable surface contamination level

5000 dpm per 100cm2 (α) averaged over an area not greater than 1 m2

Maximum acceptable surface contamination level

15,000 dpm per 100cm2 (α) applicable to an area of not more than 100 cm2

Removable surface contamination level

1,000 dpm per 100cm2 (α) applicable to an area of not more than 100 cm2

1USNRC Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material.

For uranium that exceeds the natural isotopic composition, the alpha radioactivity (due principally to U234) contributes a greater fraction of the total uranium radioactivity relative to the beta-gamma radioactivity. When measuring β-γ radioactivity for comparison to the regulatory guidelines (based on α dpm) a correction is made to account for the actual total αradioactivity. Total uranium conversion factors of 3.2 and 2.5 (Tables 3, 4) were developed and utilized for the final survey program of the former Building 2 decontamination project (Engelhard 1996). This conversion factor is detector specific for use of the GM pancake (15.5 cm2) detector and large area (425 cm2) gas flow proportional detector, respectively. Use of other instrumentation will require the determination of appropriate conversion factors. No conversion factor is required for the α removable contamination results.

Therefore:

Measured Activity (dpm/100cm2) = Net Count Rate / EFF x (Area / 100 cm2)

Total uranium (alpha) = Measured Activity x U Conversion Factor

Soil Sample Radioanalysis4.2.3High resolution gamma spectroscopy (high purity intrinsic germanium detector) was used to assess Site soil samples for uranium. This method provides a sensitive, fast, and cost effective approach to meeting project and site contamination assessment objectives. The principal advantages are that information can be derived for all the principal nuclides of uranium and most progeny; U-238 (Th-234 at 93 keV), U-235 from the U-235 isotope gammas, and U-234indirectly from the U-235 and the U235/U238 ratio (or knowledge of enrichment). The U-234 values are based on the measured U-235 concentration and a nominal U234/U235 activity ratio for this project site. A U234/U235 ratio of 22 was used and is consistent with the measured ratio from alpha spectroscopy results (Table 5). Selected samples were analyzed for isotopic uranium by alpha spectroscopy; alpha spectroscopy can directly determine U235, U234, and U238 concentrations and is useful to confirm U235 isotopic composition. It is expected that the proposed decommissioning will utilize gamma spectroscopy as the primary soil analysis (on-site or rapid turn-around laboratory service) method supplemented by isotopic uranium measurements on selected samples determined by BASF and the operations contractor.


4.3 Personnel Training and PPE

All Site personnel will require 40 hour Hazwoper training for hazardous work operations and site specific radiological training. Radiological controls personnel shall have appropriate classroom training commensurate with their work function. PPE requirements for each phase of the planned work shall be documented in SOP’s or field procedures. Radiation Work Permits (RWPs) should be used for field activities.

4.4 Radioactive Waste Management & Disposal

The work activities related to implementing the remedial action will generate radioactive waste consisting primarily of soils, debris, and contaminated materials such as PPE, etc. Waste minimization practices will be implemented to reduce the generation of radioactive waste, particularly clean items brought into the radiologically controlled area. A clear and efficient protocol for the excavation, monitoring, sampling, and segregation of soils is required due to the large volumes of soils to be evaluated. Any radioactive waste that is generated, such as contaminated personnel protective equipment or equipment not able to be decontaminated, will be properly packaged, labeled, segregated, and stored on-site pending off-site disposal.

It is expected that BASF will arrange for waste disposal as was done for the former Building 2 decommissioning radioactive waste. LLRW (low level radioactive waste) was packaged and subsequently stored in a secure Site building, pending identification of an approved off-site disposal facility for this material. This low level radioactive waste material consisted of soil, scabbling dust, piping, concrete, and other decommissioning debris contaminated with low enrichment uranium (LEU) and other contaminants (e.g. polychlorinated biphenyls [PCBs] and cadmium).

BASF completed the inspection and radiation survey with additional sampling as needed to characterize the waste packages for waste acceptance and disposal at the USEI Grand View, Idaho disposal facility. USEI then provided an assessment of the potential waste stream categorization and packaging options to meet U.S. Department of Transportation (DOT) and USEI waste acceptance and radiological activity criteria.

Waste packages had to be individually inspected and segregated by package identification numbers into one of four waste streams identified in the USEI Assessment Report as listed below:

• Soil (Non-hazardous, non-RCRA);• Soil (RCRA D006, PCBs


It is anticipated that this same process and protocols will be followed. In the event that extensive quantities of material require off-site disposal, rail haul originating from the modular transportation facility in Worcester, MA may be undertaken.

4.5 Site Security

The Site perimeter has an 8-foot high galvanized steel fence in place. Personnel and vehicle access to the Site work areas shall be controlled. No personnel will be allowed access to the Site, other than approved BASF and PM staff and Contractor personnel unless specifically authorized by BASF. Management and other personnel requiring temporary and/or infrequent Site access shall have an appropriate level of site-specific training and will be escorted during tours or inspections in the work area. All entries to the work area shall be posted and secured at end-of-shift activities.

4.6 Quality Assurance

The Program Manager (or designate) will oversee all on-site decommissioning activities and will provide internal management and QA control. He will conduct regular field audits of the decommissioning activities, and periodically participate in project and safety review meetings which review the implementation of specific QA/QC data quality objectives. The PM (or designate) will independently audit the ongoing activities and validate that decontamination and survey activities are conducted in accordance with established RCP policies and procedures.

All radiological surveys will be performed by trained and qualified personnel, and will be conducted in accordance with standard, written procedures. Each step of the decommissioning program and confirmatory surveys will be documented, to develop a stand-alone record of the decommissioning process that will be suitable for internal audit and regulatory review. Survey procedures developed for this project will be based on the guidance provided in the NRC technical document, NUREG-1575 (MARSSIM, see NRC 2000).

AU procedures and any subsequent revisions will be reviewed and approved by the decommissioning contractor management and subject to concurrence by the Program Manager and BASF. All instrumentation used in performing the surveys will be maintained in accordance with industry specifications. Calibration schedules will be established and adhered to for all instruments and analytical equipment. The instrumentation will be calibrated based on NIST-traceable standards or standards of other acceptable organizations and calibration records will be maintained. Daily tests of instrumentation will be performed against performance criteria prior to its use. BASF considers instrumentation calibration and demonstrated performance to be a significant aspect of the QA program, and care will be taken to assure that this is properly done and validated by internal audits of the program.

The data quality management program will conform to established QA procedures and the guidance provided in NUREG-1575. In-progress field procedural and measurement information will be documented and transferred in a computer generated format. Laboratory analytical data will be recorded on standard forms.

The selected analytical laboratory(s) will be required to meet radiological QA standards for sample handling and preparation, analysis, and record keeping. Field and laboratory duplicate samples at a ratio of 1 for every 10 samples will be obtained. All samples collected in the field


will be carefully packaged, labeled, and recorded. Sample custody will be maintained by a specified individual at all times with standard chain-of-custody records used to document custody from the field through laboratory analysis.

Analytical services for radiochemical and RCRA analysis of soil will be obtained from a licensed contracted laboratory. In addition, an independent, separate laboratory will be contracted to provide analysis of QA samples, which will represent a portion (approximately 10%) of the total samples. On-site gamma spectroscopy analytical services will require submittal of the contractor’s operating, calibration, and validation standard operating procedures. QA samples shall be required from a licensed off-site laboratory.

Surveillance of the effectiveness of the QA program will be validated at the project level by continuous monitoring of the work against defined procedures and standards. Scheduled and unannounced QA audits will be performed and the results recorded on checklists. Any deficiencies will be documented, and project personnel will be required to resolve these deficiencies.

4.7 Final Site Status Survey and Report

A Final Status Survey to assess the final condition of the site will be conducted after the completion of decommissioning activities. The Final Status Survey will be conducted in accordance with the guidelines provided in NUREG-1575, and all data from prior surveys used in the Survey will have been obtained under these guidelines. The results of the Final Status Survey will be documented in a Final Status Survey Report (FSR) prepared by BASF.

The RCP may choose to perform confirmatory surveys during the decommissioning process to validate the adequacy and accuracy of the decommissioning program and Final Status Survey. BASF will make available the work areas for inspection, and provide any documentation requested and collected during the decommissioning activities and site surveys.


5.0 PUBLIC INVOLVEMENT IN THE DWP

This decommissioning work plan (DWP) has already indicated BASF’s intent to decommission the Site and maintain restrictions for future use of the Site. The Owner is obligated by the RCP requirements to document in the DWP how the advice of individuals and institutions in the community who may be affected by the decommissioning has been sought and incorporated, as appropriate, following analysis of that advice. BASF will continue to work with the regulatory agencies to provide the public and local regulators with information on the proposed decommissioning work and respond to any resulting concerns. Once this DWP is issued to the public and local regulators, and feedback is obtained, that input will be incorporated into the DWP.

All requests for information and/or Site access from the public or press shall be directed to the BASF Project Manager.


6.0 REFERENCES

(AEI, 2014a) Radiological Risk Assessment and Technical Basis for the Derived Concentration Guideline Value (DCGL), AEI Consultants for BASF, February 5,, 2014.

(AEI, 2014b) Radiological Field Operations Plan, AEI Consultants for BASF, February 2014.

(Engelhard, 1995) Site Characterization Program Exterior of Plainville, Massachusetts Plant Engelhard Corporation, R. Berlin, et. al., for Engelhard Corporation., October 1995.

(Engelhard, 1996) Final Work Plan for Building Interior Decontamination at the Engelhard Corporation Site, Plainville, Massachusetts, Foster Wheeler Environmental Corporation for Engelhard Corporation, June 1996.

(Engelhard, 1997) Draft Decommissioning Work Plan for [AOC Bl7/16] Soils and Sediments at the Engelhard Corporation Plainville, Massachusetts Facility, R. Berlin, et. al., for Engelhard Corporation, June 1997.

(Engelhard, 2001) Preliminary Design Report for Soil Remediation Project (for AOC B/7), Marin Environmental, for Engelhard Corporation, July 2001.

(Engelhard, 2004) Supplemental Design Investigation (SDI) Courtyard Area (AOC B/7) Report, ECS for Engelhard Corporation, July 9, 2004.

(Environ, 2010) Final Project Report, Decommissioning of Building 11 Radioactive Material Storage Area, Environ Corporation for BASF Corporation, March, 2011.

(NRC, 1997) Letter Report-Confirmatory Survey for Survey Units 1 and 2, Interior of Building 2, Engelhard Plant, Plainville, MA (Document No. 070- 139: RFTA No.97-7), U.S. Nuclear Regulatory Commission, February 4, 1997.

(NRC, 2000) Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), NUREG-1575, U.S. Nuclear Regulatory Commission (NRC), Revision 1.

TABLES

Table 1 Results of Beta-Gamma Radiation Surveys for Soil Borings

SDI Courtyard Investigation Project Phase 1 December 2003 and Phase 2 March 2004

Survey Date

Grid Location Name Type Depth bgs

Radiation Survey Results1 (cpm) for Sampling Intervals

12/19/03 60W60N B9 slab B101 Boring 12' Background Soil Radioactivity 12/16/03 80W40N Courtyard B102 Boring 12' Background Soil Radioactivity 12/16/03 70W40N Courtyard B1032 Boring 6' (240) 0-2' bgs, (170) 2-6' bgs 12/20/03 70W40N Courtyard B1034 Boring (alternate) 6' (330) 4-6' bgs 12/23/03 60W20N B12 slab B104 Boring 16' Background Soil Radioactivity 12/18/03 60W30N B12 slab B1053 Boring 14' (260) 2-4' bgs 12/20/03 40W30N B2 slab B106 Boring 12' Background Soil Radioactivity 12/20/03 110W20N B6 slab B107 Boring 12' Background Soil Radioactivity 12/17/03 90W20N Courtyard B108 Boring 12' Background Soil Radioactivity 12/16/03 80W30N Courtyard B109 Boring 12' Background Soil Radioactivity 12/23/03 50W20N B12 slab B110 Boring 16' Background Soil Radioactivity 12/17/03 90W10N Courtyard B111 Boring 12' Background Soil Radioactivity 12/23/03 50W10N B3 slab B112 Boring 16' Background Soil Radioactivity 12/17/03 100W00N Courtyard B113 Boring 12' Background Soil Radioactivity 12/23/03 60W00N B3 slab B114 Boring 16' Background Soil Radioactivity 3/5/04 70W60N B9 slab B116 Boring 14.5' Background Soil Radioactivity 3/8/04 120W50N B6 slab B117 Boring 18' Background Soil Radioactivity 3/12/04 90W30N Courtyard B118 Boring 18' Background Soil Radioactivity 3/9/04 70W30N B12 slab B119 Boring 18' Background Soil Radioactivity 3/12/04 80W20N Courtyard B120 Boring 18' Background Soil Radioactivity 3/9/04 60W20N B12 slab B121 Boring 18' Background Soil Radioactivity 3/8/04 40W20N B2 slab B122 Boring 18' Background Soil Radioactivity 3/12/04 100W10N Courtyard B123 Boring 18' Background Soil Radioactivity 3/9/04 40W10N B2 slab B124 Boring 8' Background Soil Radioactivity 3/8/04 120W20N B6 slab B125 Boring 18' Background Soil Radioactivity 3/10/04 60W40N Courtyard B126 Boring 20' Background Soil Radioactivity 3/9/04 80W30N Courtyard B127 Boring 18' Background Soil Radioactivity 3/8/04 99W63N B9 slab B128 Boring 18' Background Soil Radioactivity 3/10/04 50W40N Courtyard B129 Boring 18' Background Soil Radioactivity 3/11/04 50W40N Courtyard B130 Boring 18' Background Soil Radioactivity 12/19/03 110W50N B6 slab MW101 Monitoring Well 14' Background Soil Radioactivity 12/19/03 70W50N B9 slab MW102 Monitoring Well 14' Background Soil Radioactivity

Table 1 Results of Beta-Gamma Radiation Surveys for Soil Borings

SDI Courtyard Investigation Project Phase 1 December 2003 and Phase 2 March 2004

12/18/03 50W30N B12 slab MW103 Monitoring Well 14' Background Soil Radioactivity 12/18/03 70W10N Courtyard MW104 Monitoring Well 14' Background Soil Radioactivity 3/5/04 70W40N Courtyard MW106 Monitoring Well 12' (1600) 0-.5' bgs, (1200) 0-2'

12/19/03 50W50N Courtyard P15 Piezometer 12' Background Soil Radioactivity 12/22/03 60W40N Courtyard P16 Piezometer 12' Background Soil Radioactivity 12/22/03 40W40N Courtyard P17 Piezometer 12' Background Soil Radioactivity 3/10/04 60W40N Courtyard PW20 Monitoring Well 18' Background Soil Radioactivity 3/10/04 50W40N Courtyard PW21 Monitoring Well 18' Background Soil Radioactivity 3/11/04 50W40N Courtyard PW23 Monitoring Well 18' Background Soil Radioactivity 12/22/03 100W50N Courtyard SS101 Shallow Boring 4' Background Soil Radioactivity 12/22/03 100W40N Courtyard SS102 Shallow Boring 4' Background Soil Radioactivity 12/20/03 40W40N Courtyard SS103 Shallow Boring 4' Background Soil Radioactivity 12/22/03 100W40N Courtyard SS104 Shallow Boring 4' Background Soil Radioactivity 3/4/04 60W40N Courtyard TP01 Test Pit 6-10' Maximum (2000) 1' bgs see note 5 3/4/04 70W40N Courtyard TP02 Test Pit 6-10' Maximum (1200) 3' bgs see note 5 3/4/04 80W40N Courtyard TP03 Test Pit 6-10' Maximum (800) 1' bgs 3/4/04 80W40N Courtyard TP04 Test Pit 6-10' Maximum (1000) 1' bgs 3/4/04 70W40N Courtyard TP05 Test Pit 6-10' Maximum (600) 1' bgs 3/4/04 60W40N Courtyard TP06 Test Pit 6-10' Maximum (600) 1' bgs 3/4/04 50W40N Courtyard TP07 Test Pit 6-10' Maximum (600) 1' bgs 3/4/04 60W40N Courtyard TP08 Test Pit 6-10' Background Soil Radioactivity 3/4/04 50W40N Courtyard TP09 Test Pit 6-10' Background Soil Radioactivity 3/4/04 50W40N Courtyard TP10 Test Pit 6-10' Background Soil Radioactivity 3/4/04 50W40N Courtyard TP10A Test Pit 6-10' Background Soil Radioactivity 3/4/04 50W40N Courtyard TP11 Test Pit 6-10' Background Soil Radioactivity 3/4/04 50W40N Courtyard TP12 Test Pit 6-10' Background Soil Radioactivity

1. Measurements for elevated soil radioactivity in cpm - measured with GM pancake detector. 2. Cuttings from this location B103 estimated at 4-6' bgs measured 500-600 cpm. 3. Cuttings from this location B105 estimated at 4-8' bgs measured maximum of 250 cpm. 4. Alternate boring location for B103 approximately 2 meters east of original boring location. 5. Uranium isotopic analyses for soil samples (GEL Laboratories).

TABLE 2

URANIUM CONCENTRATIONS IN SOIL BASF SITE, PLAINVILLE, MA

ALARA (

TABLE 3

Summary of Planned Radiological Remediation and Investigation Activities

Uranium 238 Uranium 235 Uranium 234SAMPLE Th234 (93 keV) U235 (186 keV) U235 x 22

AREA IDENTIFICATION Conc +/- MDA Conc +/- MDA Conc +/- MDA BGS SUM DCGLpCi/g pCi/g pCi/g pCi/g pCi/g pCi/g pCi/g pCi/g pCi/g code FRACTION TEST PLANNED RADIOLOGICAL ACTION

1 UCY EP-SBRD013-1-.5-2 15.2 2.0 1.8 4.5 0.2 0.1 99.4 4.3 2.0 A 0.5 Remediate with Bldg 3 slab removal2 UCY EP-SBRD041-1-.5-2 21.6 1.5 1.3 4.3 0.3 0.1 94.0 5.8 1.9 A 0.6 NaI investigation or remediation3 UCY EP-SBRD108 6.9 0.7 0.9 1.6 0.1 0.1 35.4 2.9 1.7 A 0.2 Remediate with Bldg 3 slab removal4 UCY RAD01(SDI 3/4/04) 881.0 30.4 37.4 6.3 432.0 21.3 B 11.4 >DCGL Remediation5 UCY RAD02 (SDI 3/4/04) 195.0 6.1 7.4 1.2 63.3 3.5 A 2.4 >DCGL Remediation6 SED EP-SDRD05 34.7 1.7 1.4 5.3 0.3 0.1 115.7 6.3 2.0 A 0.8 Remediate with sediment removal7 SED EP-SDRD05A-.5-1 9.0 1.0 1.1 2.8 0.5 0.1 60.6 11.7 1.5 A 0.3 Remediate with sediment removal8 LF/DW EP-SBRD026 8.9 1.4 1.5 1.5 0.3 0.1 32.7 6.9 2.0 A 0.2 NaI investigation or remediation9 LF/DW EP-SBRD030-7-12-14 243.5 4.5 3.2 8.7 1.7 0.1 192.0 36.7 3.2 G 3.2 >DCGL > 4'bgs remediate if practicable with Dry Well removal

10 LF/DW EP-SBRD033-0-.5 34.1 2.3 1.9 13.7 0.3 0.1 301.5 7.0 2.3 A 1.5 >DCGL Remediate with Dry Well removal11 LF/DW EP-SBRD030-6-10-12 103.7 3.0 2.1 3.7 0.7 0.1 80.6 15.7 2.3 F 1.4 >DCGL > 4'bgs remediate if practicable with Dry Well removal12 LF/DW EP-SBRD033A-2-10-12 78.2 3.0 2.1 3.7 0.7 0.1 81.5 15.8 2.2 F 1.1 >DCGL > 4'bgs remediate if practicable with Dry Well removal13 LCY EP-SBRD002-1-.5-2 9.3 1.2 1.3 1.7 0.4 0.1 37.4 7.7 2.0 A 0.2 NaI investigation or remediation14 LCY EP-SSRD009 15.6 2.3 2.1 4.2 0.2 0.1 93.3 4.9 2.9 A 0.5 NaI investigation or remediation15 LCY EP-SSRD010 18.8 2.3 1.9 5.8 0.2 0.1 127.7 4.8 2.4 A 0.7 NaI investigation or remediation16 LCY EP-SBRD010-1-.5-2 7.7 1.0 1.3 1.6 0.1 0.1 34.4 2.6 2.0 A 0.2 NaI investigation or remediation17 LCY EP-SSRD034 10.3 1.3 1.4 1.5 0.3 0.1 32.6 6.8 1.8 A 0.2 NaI investigation or remediation18 LCY EP-SSRD039 30.0 2.6 2.1 3.6 0.2 0.1 78.7 4.7 3.0 A 0.6 NaI investigation or remediation19 LCY EP-SBRD039-1-.5-2 20.7 1.7 1.4 3.5 0.2 0.1 77.1 5.0 1.9 A 0.5 NaI investigation or remediation20 LCY EP-SSRD107 6.2 0.8 1.0 1.1 0.1 0.1 24.3 2.7 1.8 A 0.2 NaI investigation or remediation21 LCY EP-SBRD107-1-.5-2 16.8 1.8 1.5 2.9 0.2 0.1 63.5 4.8 1.9 A 0.4 NaI investigation or remediation22 LCY EP-SBRD107-2-2-4 7.2 0.7 0.9 1.2 0.1 0.1 25.3 2.3 1.7 B 0.2 NaI investigation or remediation23 LCY EP-SBRD125-1-.5-2 8.9 1.0 1.1 2.0 0.4 0.1 43.9 8.6 1.6 A 0.3 NaI investigation or remediation24 LCY EP-SBRD126-1-.5-2 7.3 0.7 0.9 1.5 0.3 0.1 32.1 6.4 1.4 A 0.2 NaI investigation or remediation25 LCY EP-SBRD126-2-2-4 13.4 0.9 0.9 2.1 0.4 0.1 46.8 9.0 1.4 B 0.3 NaI investigation or remediation26 LCY EP-SSRD130 6.5 0.7 0.9 1.1 0.2 0.1 24.6 4.7 1.6 A 0.2 NaI investigation or remediation27 LCY EP-SBRD130-1-.5-2 8.2 0.8 0.9 1.3 0.3 0.1 28.8 5.5 2.0 A 0.2 NaI investigation or remediation28 B2SUB TUNNEL/TRENCH(NRC#10) 77.7 7.2 190.1 7.4 4182.2 A 16.5 >DCGL Remediate with Bldg 2 slab removal29 B2SUB TUNNEL/TRENCH(NRC #9) 656.0 34.0 101.4 4.4 2230.8 A 15.1 >DCGL Remediate with Bldg 2 slab removal30 B2SUB TUNNEL/TRENCH(NRC#8) 76.1 6.2 57.2 2.5 1258.4 A 5.5 >DCGL Remediate with Bldg 2 slab removal31 B2SUB TUNNEL4(ENG-NRC#10) 137.7 23.9 525.8 A 3.4 >DCGL Remediate with Bldg 2 slab removal32 B2SUB 2M/NPITWALL(NRC#5) 29.4 2.1 18.5 0.8 407.0 A 1.8 >DCGL Remediate with Bldg 2 slab removal33 B2SUB TUNNEL3 (ENG-NRC#9) 14.8 15.7 345.4 A 1.4 >DCGL Remediate with Bldg 2 slab removal34 B2SUB TUNNEL2 (ENG-NRC#8) 3.7 13.6 298.3 A 1.2 >DCGL Remediate with Bldg 2 slab removal35 B2SUB 2M/PITBOTTOM(NRC#7) 8.1 1.0 3.3 0.3 72.6 A 0.4 Remediate with Bldg 2 slab removal

Page 1 of 1

Table 4: Dependence on the ratio of total to measured radioactivity for various detector types.Determination of Conversion FactorsMeasured dpm to total uranium (alpha) dpm

3.7 wt % U23522 U234/U235 activity ratio

6.25E-03 specific activity (Ci/g) U234 0.028 wt% U2342.16E-06 specific activity (Ci/g) U235 3.70 wt% U2353.36E-07 specific activity (Ci/g) U238 96.27 wt% U238

Based on 100 grams:Grams Ci U234 U235 Th231 U238 Th234 Pa234m Total

U234 0.0282 1.76E-04 3.91E+08 3.91E+08U235 3.70 8.00E-06 1.78E+07 1.78E+07 3.55E+07U238 96.2718 3.24E-05 7.19E+07 7.19E+07 7.19E+07 2.16E+08

Total 6.42E+08Total beta dpm = 1.62E+08Total alpha dpm = 4.80E+08alpha to beta ratio = 3.0

Detector Detector Thin Window GM pancake Large Area Gas Flow

calc dpm Bavg Efficiency cpm Efficiency cpmU234 3.91E+08 alpha 0.028 1.07E+07 0.038 1.47E+07U235 1.78E+07 alpha 0.028 4.88E+05 0.038 6.66E+05Th231 1.78E+07 0.079 Tc99 .097 Bavg 0.150 2.66E+06 0.100 1.78E+06U238 7.19E+07 alpha 0.028 2.01E+06 0.038 2.70E+06Th234 7.19E+07 0.050 0.100 7.19E+06 0.070 5.03E+06Pa234m 7.19E+07 0.825 Y90 0.76 Bavg 0.290 2.08E+07 0.200 1.44E+07Total 6.42E+08 Total 4.39E+07 Total 3.92E+07

Page 1 of 2

Table 4: Dependence on the ratio of total to measured radioactivity for various detector types.

Reference Eff Measured dpm Total U Conversion Factor Total U Conversion Factor

GM Detector 0.29 1.52E+08 3.2Large Area Gas Flow 0.20 1.96E+08 2.5

Page 2 of 2

BASF SITEPLAINVILLE, MA.

TABLE 5

LABORATORY INTERCOMPARISON OF URANIUM ISOTOPIC SOIL ANALYSES GAMMA SPECTROSCOPY(G) AND ALPHA SPECTROSCOPY(A) RESULTS

U238 U238 U235 U235 U234 U234 U235 wt% U Total U Total U4/U5G A G A G calc A G calc A calc G A A

SAMPLE ID pCi/g pCi/g pCi/g pCi/g pCi/g pCi/g pCi/g pCi/g ratio

1 EP-SSRD33 34.1 27.7 13.70 9.98 301 197 5.9 5.3 349 234 19.72 EP-SSRD13 2.6 3.2 0.86 1.07 18.9 19.0 4.8 4.9 22.4 23.3 17.83 EP SDRD05A-.5-1 9.0 8.8 2.75 2.79 60.6 49.3 4.5 4.7 72.4 60.9 17.74 FW-SM-COMP 122 154 29.3 38.2 645 768 3.6 3.7 796 961 20.15 EP SSRD06 2.9 3.3 0.68 0.62 14.9 13.7 3.5 2.8 18.4 17.7 22.16 EP-SBRD39-2-2-4 4.8 7.8 0.97 1.12 21.4 26.4 3.1 2.2 27.2 35.3 23.57 EP SBRD126-1-.5-2 7.3 7.8 1.46 1.34 32.1 25.8 3.0 2.6 40.8 35.0 19.38 EP SBRD02-1-.5-2 9.3 7.2 1.70 1.37 37.4 26.4 2.8 2.9 48.4 34.9 19.29 EP SSRD124 4.2 4.7 0.73 1.01 16.0 16.8 2.7 3.2 20.9 22.6 16.710 EP SBRD12-2-2-4 4.0 6.9 0.68 0.71 15.0 12.6 2.6 1.6 19.7 20.1 17.711 EP SSSB26-0-.5 8.9 8.5 1.48 1.12 32.7 22.4 2.5 2.0 43.0 32.0 20.012 EP-SBRD07-1-.5-2 4.3 4.2 0.68 0.58 15.0 11.5 2.4 2.1 19.9 16.3 19.813 EP-RF1C 5.0 6.8 0.80 0.81 17.6 17.4 2.4 1.8 23.3 24.9 21.414 EP-SDRD05 34.7 33.1 5.26 3.34 116 89.1 2.3 1.5 156 126 26.715 EP-SBRD130-2-2-4 6.5 8.4 0.95 1.12 20.8 19.4 2.2 2.0 28.3 28.9 17.316 EP SSRD34 10.3 9.7 1.48 1.54 32.6 29.5 2.2 2.4 44.4 40.7 19.117 EP SDRD05C-0-.5 5.1 4.2 0.72 0.47 15.8 11.2 2.2 1.7 21.6 15.8 23.718 EP-SBRD130-3-4-6 5.6 5.4 0.74 0.61 16.4 11.0 2.0 1.7 22.7 17.0 18.019 EP-SSRD24 6.0 4.4 0.79 0.66 17.3 13.7 2.0 2.3 24.1 18.8 20.820 EP-SSRD05 4.4 6.4 0.56 0.93 12.3 14.8 1.9 2.2 17.3 22.1 15.921 EP SSSB21-0-.5 4.8 5.1 0.53 0.55 11.6 10.5 1.7 1.6 16.9 16.2 19.223 EP SBRD10-2-2-4 2.8 1.8 0.29 0.25 6.3 5.3 1.6 2.1 9.4 7.3 21.124 EP-SBRD19-6-10-12 7.1 7.2 0.67 0.60 14.7 12.6 1.4 1.3 22.4 20.4 20.925 EP SBRD115-4-6-8 15.3 11.0 1.38 1.01 30.3 19.1 1.4 1.4 47.0 31.2 18.926 EP SBRD32-2-10-12 14.4 16.0 0.89 0.99 19.6 16.8 1.0 1.0 35.0 33.8 17.027 EP-SBRD128-7-12-14 20.5 20.6 0.93 1.06 20.5 17.0 0.70 0.80 42.0 38.6 16.028 EP-SBRD27-6-10-12 10.4 14.1 0.45 0.68 10.0 11.7 0.68 0.75 20.8 26.4 17.229 EP-SBRD26-7-12-14 10.4 8.5 0.39 0.43 8.6 7.2 0.58 0.78 19.4 16.2 16.830 EP SBRD30-7-12-14 243 458 8.73 20.76 192 275 0.55 0.70 444 753 13.2

Correlation Coefficient 0.99 0.95 0.98 0.94 0.97

(avg + 1sigma) wt% U235 (A) 3.4 (avg + 1sigma) U4/U5 21.9

Page 1 of 1

FIGURES

LEGEND:

70N, 120W70N, 30W

0N, 0W

B-117

B-125

0N, 30W

60N, 30W

0N, 120W

B-107

MW-101

60N, 120W

B-116B-128

FORMER BLDG 12

FORMER BLDG 3

COURTYARD AREA

OBMW-28

OBMW-06FORMER TRANSFORMER YARD

B-113

B-111

B-110

B-112

B-114

SS-101

SS-102

B-105

P-16

MW-103

PW-15

B-101

MW-102

B-106

SS-103

40W

60W

80W100W

B-121

B-122

B-124

B-126

B-129

PW-17PW-20

PW-23B-130

PW-21

SB-RD33

MW-18OBMW-05

MW-25B

OBMW-25A

DRY WELL

CB

(SS-105)DW

SS-104MW-27

B-109

B-119

B-102

MW-106B-103

(SS-106)

B-108

MW-104

B-104

B-123

B-120

B-118

B-127

SD-18

SD-19

SD-21

SD-20

SD-17

SD-16

SD-06

SD-28

SD-24

SD-26

SD-27

SD-23

SD-43

MW-105

SB-RD03

SB-RD101

SB-RD102

SB-RD103

SB-RD04

SB-RD127

SB-RD104 SB-RD39

SB-RD09

SB-RD113 SB-RD14

SB-RD06 SB-RD108

SB-RD13

SB-RD108

SB-RD41

SB-RD106

SB-RD08 SB-RD111SB-RD12

SB-RD05

SB-RD40

SS-RD43SS-RD37

SS-RD35

SS-RD36

SB-RD33A

SB-RD131

SB-RD132SB-RD130

SB-RD01SB-RD129

SB-RD126

SB-RD125

SB-RD02SB-RD105

SB-RD124 SB-RD123 SB-RD122

SB-RD07



SB-RD10SB-RD112

SB-RD18 SB-RD121

SB-RD25

SB-RD19

SB-RD114

SB-RD23

SB-RD115SB-RD115A

SB-RD24

SB-RD27


SB-RD22

SB-RD26

SB-RD21

SB-RD20

SB-RD28

SB-RD29

SB-RD30SB-RD117

SB-RD32

SB-RD120SB-RD128

SB-RD116SS-RD78

SS-RD77

SS-RD79

SS-RD78

SS-RD80

SS-RD81 SS-RD82

SS-RD34

SS-RD42

SS-RD45

SS-RD47

SS-RD48SS-RD50

SS-RD51SS-RD52

SS-RD53 SS-RD54 SS-RD55

SS-RD47

RAD-02RAD-01

SD-RD02SD-RD03

SD-RD04

SD-RD09

SD-RD05

SD-RD05BSD-RD05A

SD-RD05C

SD-RD05DSD-RD05E

SD-RD06

SD-RD07

SD-RD08

SS-RD38 SS-RD46

SS-RD32

SS-RD31

SS-RD29

SS-RD30

SS-RD27

SS-RD25

SS-RD24

SS-RD23

SS-RD49

2H

2K

2B2A

2C

2N

2M 2L

TUNNEL

2E

2D

2G

2F

2I

2J

CHEM LAB

2P

LEACH PITS

SHORELINE TURNPIKE LAKE

PLAINVILLE, MASSACHUSETTS

AOC B, 7, and 16

BASF CORPORATION SITE

FIGURE 1A

GRID LAYOUT/SPACING IS 10m X 10m

BEDROCK MONITORING WELL

SDI SOIL BORINGS

OVERBURDEN MONITORING WELL/PIEZOMETER

SHALLOW SOIL SAMPLING LOCATION

SEDIMENT SAMPLING LOCATION

SHALLOW SOIL RADIONUCLIDE SAMPLING LOCATION

RADIONUCLIDE SOIL BORING

SEDIMENT RADIONUCLIDE SAMPLING LOCATION

ALARA INVESTIGATION AREAS

RAD SAMPLE LOCATION EXCEEDING DESIGN DOSE VALUE

RAD EXCAVATION AREA PROPOSED 0-2' EXCAVATION ANDALARA REMEDIATION SEDIMENTS

ALARA INVESTIGATION

SUB-SLAB SOILS INVESTIGATION

SUB-SURFACE LEACHFIELD PIPING INVESTIGATION

LEGEND:

70N, 120W70N, 30W

0N, 0W

B-117

B-125

0N, 30W

60N, 30W

0N, 120W

B-107

MW-101

60N, 120W

B-116B-128

FORMER BLDG 12

FORMER BLDG 3

COURTYARD AREA

OBMW-28

OBMW-06FORMER TRANSFORMER YARD

B-113

B-111

B-110

B-112

B-114

SS-101

SS-102

B-105

P-16

MW-103

PW-15

B-101

MW-102

B-106

SS-103

40W

60W

80W100W

B-121

B-122

B-124

B-126

B-129

PW-17PW-20

PW-23B-130

PW-21

SB-RD33

MW-18OBMW-05

MW-25B

OBMW-25A

DRY WELL

CB

(SS-105)DW

SS-104MW-27

B-109

B-119

B-102

MW-106B-103

(SS-106)

B-108

MW-104

B-104

B-123

B-120

B-118

B-127

SD-18

SD-19

SD-21

SD-20

SD-17

SD-16

SD-06

SD-28

SD-24

SD-26

SD-27

SD-23

SD-43

MW-105

SB-RD03

SB-RD101

SB-RD102

SB-RD103

SB-RD04

SB-RD127

SB-RD104 SB-RD39

SB-RD09

SB-RD113 SB-RD14

SB-RD06 SB-RD108

SB-RD13

SB-RD108

SB-RD41

SB-RD106


SB-RD05

SB-RD40

SS-RD43SS-RD37

SS-RD35

SS-RD36

SB-RD33A

SB-RD131

SB-RD132SB-RD130

SB-RD01SB-RD129

SB-RD126

SB-RD125

SB-RD02SB-RD105


SB-RD07



SB-RD10SB-RD112

SB-RD18 SB-RD121

SB-RD25

SB-RD19

SB-RD114

SB-RD23

SB-RD115SB-RD115A

SB-RD24

SB-RD27


SB-RD22

SB-RD26

SB-RD21

SB-RD20

SB-RD28

SB-RD29

SB-RD30SB-RD117

SB-RD32

SB-RD120SB-RD128

SB-RD116SS-RD78

SS-RD77

SS-RD79

SS-RD78

SS-RD80

SS-RD81 SS-RD82

SS-RD34

SS-RD42

SS-RD45

SS-RD47

SS-RD48SS-RD50

SS-RD51SS-RD52

SS-RD53 SS-RD54 SS-RD55

SS-RD47

RAD-02RAD-01

SD-RD02SD-RD03

SD-RD04

SD-RD09

SD-RD05

SD-RD05BSD-RD05A

SD-RD05C

SD-RD05DSD-RD05E

SD-RD06

SD-RD07

SD-RD08

SS-RD38 SS-RD46

SS-RD32

SS-RD31

SS-RD29

SS-RD30

SS-RD27

SS-RD25

SS-RD24

SS-RD23

SS-RD49

2H

2K

2B2A

2C

2N

2M 2L

TUNNEL

2E

2D

2G

2F

2I

2J

CHEM LAB

2P

LEACH PITS

SHORELINE TURNPIKE LAKE


AOC B, 7, and 16


FIGURE 1B



SDI SOIL BORINGS

OVERBURDEN MONITORING WELL/PIEZOMETER

SHALLOW SOIL SAMPLING LOCATION

SEDIMENT SAMPLING LOCATION

SHALLOW SOIL RADIONUCLIDE SAMPLING LOCATION

RADIONUCLIDE SOIL BORING

SEDIMENT RADIONUCLIDE SAMPLING LOCATION

ALARA INVESTIGATION AREAS

RAD SAMPLE LOCATION EXCEEDING DESIGN DOSE VALUE

RAD EXCAVATION AREA

PROPOSED 2-4' EXCAVATION ANDALARA REMEDIATION SEDIMENTS

ALARA INVESTIGATION

SUB-SLAB SOILS INVESTIGATION

SUB-SURFACE LEACHFIELD PIPING INVESTIGATION

LEGEND:


AOC B, 7, and 16


FIGURE 2

HISTORICAL ROOM 2MAND TUNNEL SOIL REMOVAL AREAS

Room 2M Tunnel, West

Tunnel, Center

Tunnel, East

Source: Decommissioning Work Plan for Soils andSediments at the Engelhard Corporation Plainville,Massachusetts Facility, R. Berlin, et. al., for EngelhardCorporation, June 9 1997.

Location Uranium Concentration (pCi/g)

U-238 U-235 Total Uraniuma #1 Room 2A 4.9 ± 0.8b 0.8 ± 0.2 22.5

#2 Room 2K 1.6 ± .06 0.2 ± 0.1 6.0

#3 Room 2L 1.3 ± 0.6

LEGEND:


AOC B, 7, and 16


FIGURE 3



SDI SOIL BORINGS

OVERBU

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