ML022970071

Maine Yankee'S License Termination Plan, Section 2, Table of Contents - Attachment 2I, Table 2I-2

ML022970071
Person / Time
Site:	Maine Yankee
Issue date:	10/15/2002
From:	Maine Yankee Atomic Power Co
To:	NRC/FSME
References
+sisprbs20060109, -nr, -RFPFR
Download: ML022970071 (251)

Text

MYAPC License Termination Plan Revision 3 October 15, 2002 MAINE YANKEE LTP SECTION 2 SITE CHARACTERIZATION

MYAPC License Termination Plan Page 2-i Revision 3 October 15, 2002 TABLE OF CONTENTS 2.0 SITE CHARACTERIZATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.2 Historical Site Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 2.2.1 Historical Data Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3 2.2.2 Decommissioning File 10 CFR 50.75(g) . . . . . . . . . . . . . . . . . . . . . . . 2-4 2.2.3 10 CFR 20.302 Submittal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-5 2.2.4 Historical Radiological Status Including Original Shutdown Status . . . 2-6 2.2.5 Current Radiological Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-7 2.2.6 Hazardous and Chemical Material Contamination . . . . . . . . . . . . . . . . 2-8 2.3 Site Characterization Survey Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-9 2.3.1 Organization and Responsibilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-9 2.3.2 Characterization Data Categories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-10 2.3.3 Characterization Survey Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-12 2.3.4 Instrumentation and Minimum Detectable Concentrations (MDCs)

Instrument Selection and Use . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-13 2.3.5 Quality Assurance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-17 2.3.6 Data Quality Objectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-20 2.3.7 Survey Findings And Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-21 2.4 Summary of Initial Characterization Survey (ICS) Results . . . . . . . . . . . . . . . 2-30 l 2.4.1 Group A Affected Structures and Surfaces . . . . . . . . . . . . . . . . . . . 2-30 2.4.2 Group B Unaffected Structures and Surfaces . . . . . . . . . . . . . . . . . . 2-31 2.4.3 Group C Affected Plant Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-31 2.4.4 Group D Unaffected Plant Systems Including the Sewage Treatment System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-32 2.4.5 Group R Environs Affected and Unaffected . . . . . . . . . . . . . . . . . . 2-33 2.4.6 Ventilation Ducts and Drains . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-34 2.4.7 Buried and Embedded Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-34 2.4.8 Asphalt, Gravel and Concrete . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-40 2.4.9 Paved Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-41 2.4.10 Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-41 2.4.11 Structures, Systems and Environs Surveyed For Hazardous Material (Groups E and H) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-42 2.4.12 Surface and Groundwater . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-42 2.4.13 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-42 2.4.14 Waste Volumes and Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-45

MYAPC License Termination Plan Page 2-ii Revision 3 October 15, 2002 2.5 Continuing Characterization (CCS) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-45 l 2.5.1 Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-46 2.5.2 Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-46 2.5.3 Nuclide Profile . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-52 2.5.4 Background Determination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-63 2.6 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-65 2.6.1 Impact Of Characterization Data On Decontamination And Decommissioning . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-65 2.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-66 ATTACHMENT 2A Non-Impacted Area Assessment ATTACHMENT 2B Characterization Data ATTACHMENT 2C Summary of Continued Characterization Data ATTACHMENT 2D Maine Yankee Site Characterization Locations of Radiological Survey Packages ATTACHMENT 2E Site and Survey Area Maps ATTACHMENT 2F Analysis of Concrete Sample Variance ATTACHMENT 2G l l

Supplemental Information Regarding Concrete Core Data Use l

MYAPC License Termination Plan Page 2-iii Revision 3 October 15, 2002 ATTACHMENT 2H l l

Forebay and Diffuser Characterization Discussion l ATTACHMENT 2I l l

Soil Sampling and Nuclide Fraction l l

List of Tables Table 2-1 Significant Soil Contamination Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-5 Table 2-2 Volumetric MDCs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-15 Table 2-3 Theoretical Scanning Sensitivities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-16 Table 2-4 Summary of ICS Material Backgrounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-43 Table 2-5 Summary of ICS Environs Background Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-44 Table 2-6 Summary of Miscellaneous Background Survey Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-45 Table 2-7 Nuclide Fractions, Contaminated Concrete Surfaces (Balance of Plant Areas) . . . . . . . . . 2-54 l Table 2-8 Nuclide Fractions for Contaminated Concrete Surfaces Special Areas . . . . . . . . . . . . . . . 2-55 l Table 2-9 Activated Concrete Nuclide Fractions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-57 Table 2-10 Activated Concrete: Deep Core Sample Activity Profile . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-58

MYAPC License Termination Plan Page 2-iv Revision 3 October 15, 2002 Table 2-11 Soil Nuclide Fractions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-59 Table 2-12 Ground and Surface Water Nuclide Fraction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-61 Table 2-13 Forebay/Diffuser Material Nuclide Fractions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-62 Table 2-14 Structural Material Backgrounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-64

MYAPC License Termination Plan Page 2-1 Revision 3 October 15, 2002 2.0 SITE CHARACTERIZATION 2.1 Overview The radiological and chemical characterization of the Maine Yankee (MY) site has been going on since pre-operational sampling was begun in 1970. Initial site characterization for decommissioning was begun in the fall of 1997 and ran through the spring of 1998.

Historical information, including the 10 CFR 50.75(g) file, employee interviews, Radiological Incident files, pre-operational survey data, spill reports, special surveys (e.g., site aerial surveys, marine fauna and sediment surveys), operational survey records and Annual Radiological Environmental Reports (including sampling of air, groundwater, estuary water, milk, invertebrates, fish and surface vegetation) to the NRC were reviewed and compiled into the Historical Site Assessment (HSA). Using the information collected during the HSA, an overall characterization plan was developed to collect measurements l and samples from plant structures, systems and open land areas to cover the areas where contamination existed, remained or had the potential to exist.

The information collected during all phases of site characterization, including the HSA, l was used during decommissioning planning to achieve the following objectives:

• Determine the radiological status of the site and facility to include identification of systems, structures, soils and water sources in which contamination exists;

• Identify the location and extent of any contamination outside the radiological restricted areas (RA);

• Estimate the source term and radionuclide mixture to support decommissioning cost estimation and decision-making for remediation, dismantlement and radioactive waste disposal activities;

• Select the instrumentation used for surveys and develop the quality assurance methods applied to sample collection and analysis;

• Perform dose assessment and FSS design; and

• Ensure the Radiation Protection Program addresses any unique radiological health and safety issues associated with decommissioning.

The initial site characterization process focused on four areas, providing both shutdown l and current data for structures, systems, radiological environs and hazardous materials environs. The extent and range of contamination were reported for structures, systems, drains, vents, embedded piping, paved areas, water and soils. In addition, activation analyses were performed on key components within the restricted area to estimate radioactive waste volumes and classes.

MYAPC License Termination Plan Page 2-2 Revision 3 October 15, 2002 The initial characterization results (ICS1) were provided to MY in the Characterization l Survey Report for the MY Atomic Power Plant, developed by GTS Duratek. After l review of this initial characterization report, it was determined that additional sampling l was needed to fully define the extent of contamination in some outdoor areas and some systems in order to design the FSS, perform dose assessments and address questions related to waste volumes. This additional sampling, which is generally referred to as l Continuing Characterization Surveys (CCS), is discussed in Section 2.5. As additional l data is required (such as concrete cores, forebay sediment, etc.), characterization samples l will be obtained; thus, CCS is an ongoing activity and is included as part of the FSS l process. l This section summarizes the key findings of the HSA and characterization survey results, as supplemented by continuing characterization. The initial characterization report and l the detailed results of continuing characterization, are maintained at the MY site and are l available for NRC review. Data from the CCS effort, due to its ongoing nature, is filed l with the appropriate characterization package associated with the system, structure, l component, or area being surveyed (or sampled). These packages are maintained in the l Plant Technical File System. The level of detail provided in this summary demonstrates l that the overall characterization plan objectives listed above have been met. In addition, l the characterization data provided in this section are consistent with NRC guidance contained in Regulatory Guide 1.179, Standard Format and Content of License Termination Plans for Nuclear Power Reactors, and sufficient to meet the review criteria set forth in NUREG-1700, Standard Review Plan for Evaluating Nuclear Power Reactor License Termination Plans.

2.2 Historical Site Assessment The Radiation Protection organization amassed tens of thousands of survey records documenting general area and component-specific radiation levels, contamination levels, system activity levels and airborne radioactivity levels during 25 years of plant operation.

These survey records reflected radiological conditions on site with frequency and detail dependent on the magnitude of radiation and contamination present in an area and the frequency with which the area was entered by the operating staff. Plant document files contained records of spill and event reports (Operations Department Unusual Occurrence Reports and Radiation Protection Department Radiological Incident Reports) as well as the required annual or semiannual radiological effluent reports to the NRC which documented any unplanned releases.

In order to ensure a complete discovery of events involving spills, leaks or other 1 "

ICS, as used in the LTP refers to the initial characterization performed by GTS Duratek, as l documented in the Characterization Survey Report for the MYAPP, 1998. It may also simply be l referred to as the GTS Duratek report. l

MYAPC License Termination Plan Page 2-3 Revision 3 October 15, 2002 operational occurrences which might have an effect on the radiological and chemical status of the site, MY also interviewed terminating employees for any recollection of such events.

2.2.1 Historical Data Review Historical records contained in the radiation protection files, 10 CFR 50.75(g) file, Annual Radiological Environmental Reports to the NRC, miscellaneous environmental reports, and one 10 CFR 20.302 submittal were reviewed to determine the location and extent of leaks and spills on site. The pertinent results of the record reviews, Initial Site Characterization surveys, and employee interviews were captured in the Historical Site Assessment (HSA). The HSA, as l supplemented, is a compilation of the approximately 140 potential events l occurring over the 25 year operating history of the plant. About two thirds of these events were potential radiological issues with the other one third being chemical or hazardous material events.

Key items identified in the HSA include:

1. Contaminated soil between the RA and Forebay, from RWST leaks;

2. Contaminated soil after the removal of a low level waste storage area (Wiscasset wall);

3. Location of a silt spreading area/construction debris landfill;

4. A waste neutralization tank drain line leak;

5. A PCC leak in the alley way;

6. Contaminated soil on Bailey Point, south of the Industrial Area (IA) trailer park, in an area where contaminated soil from the PCC leak had been stored;

7. Discrete particles throughout plant from reactor core barrel machining;

8. Contaminated soil in the ISFSI area, formerly known as the contractor parking lot;

9. A discrete particle outside warehouse 2;

10. Contaminated sumps and floor trenches in the turbine hall;

11. RA sink and decon shower drains go to sewage treatment plant;

12. Contaminated sediment in the Forebay;

13. Previous abandonment of an underground ferrous sulfide tank;

14. Snow from RA placed in ball field;

15. Contaminated soil from BWST leaks;

16. Contaminated soils in the IA trailer park; and

17. Very low levels of detectable residual radioactivity on Foxbird Island, RCA building roof, Equipment Hatch pit, and on the concrete block in the ball field dugouts.

18. Two large volume spills in the Containment Spray Building l

MYAPC License Termination Plan Page 2-4 Revision 3 October 15, 2002 None of the event records in the HSA indicated the uncontrolled release of radioactive material affecting the site beyond Bailey Point (i.e., south of Ferry Road and east of Bailey Cove).

2.2.2 Decommissioning File 10 CFR 50.75(g)

Even though MY was in operation well before the requirement to maintain a decommissioning file, the 50.75(g) file contained documentation of three areas of soil contamination and one record of a 10 CFR 20.302 submittal for burial in place of residual soil activity. The information in the decommissioning file was added to the HSA so that the affected areas could be properly addressed during site characterization.

The 50.75(g) file documented soils outside the Spray, Containment and Fuel Buildings (see Table 2-1) that were known to contain contamination from an RWST manway leak, a series of RWST siphon heater leaks, SCC/PCC leaks, as well as the storage of radioactive waste awaiting shipment in an outside, shielded storage location. Some work was also performed on contaminated components within tented enclosures located outside the RCA Storage Building which also contributed to soil and pavement contamination.

MYAPC License Termination Plan Page 2-5 Revision 3 October 15, 2002 Table 2-1 Significant Soil Contamination Events Event Date Location Volume Disposition Estimated Residual Activity RWST siphon 2/23/88 Area south and 8200 ft3 Remediated 600 6 mCi heater leak west of RWST ft3. 7600 ft3 left in place under 10 CFR 20.302.2 Removal of Low 7/92 Outside the 2000 ft3 Residual 5.9 mCi Level Waste RCA Storage contamination Storage Area Bldg and west to evaluated and high rad bunker entered into 50.75(g) file.

Silt spreading 1992, Land adjacent to 1250 ft3 Residual 12 µCi area 1993 and south of contamination Outages ballfield. evaluated and entered into 50.75(g) file.

2.2.3 10 CFR 20.302 Submittal (reference Table 2-1 above)

MY applied to the NRC on 11/2/88 (MN 88-107) to allow residual soil contamination to remain in place under the provisions of 10 CFR 20.302. The NRC approved the submittal on 8/31/89. This data is included to provide a complete historical basis for the overall site characterization. The details of the l soil contamination are presented below.

In 1988 a small outdoor leak at the inlet flange connection between the RWST siphon heater return line and an isolation valve was discovered and subsequently contained. The actual time that the leak started and the volume of water lost could not be determined. Surveys of the area adjacent to the RWST indicated ground contamination as high as 7E-3 µCi/g of Cs-137.

The leak was repaired, and the contaminated soil was removed from the area and disposed of as radioactive waste. Sample analysis of the soil removed from the area of remediation also indicated the presence of Cs-134, Sb-125 and Co-60 in addition to the Cs-137. The level of activity of these additional nuclides was 2

10 CFR 20.302 has been superceded by 10 CFR 20.2002

MYAPC License Termination Plan Page 2-6 Revision 3 October 15, 2002 approximately two orders of magnitude less than the Cs-137. Soil was excavated to a level of two to five feet below grade until the average residual Cs-137 activity had decreased to an equivalent MPC value in water of about 2E-5 µCi/ml.

Approximately 600 cubic feet of radioactive waste was generated from the excavation. Residual activity of Cs-137 in an estimated 7600 cubic feet of l remaining affected soil was 6 mCi. The location of this contaminated soil is well known and the need for further remediation will be evaluated, via sampling and analysis, during decommissioning to ensure compliance with the unrestricted use criterion. Section 5.5.1.b presents a discussion of deep soil contamination sampling in and near the RWST spill area.

2.2.4 Historical Radiological Status Including Original Shutdown Status MY ran for approximately 16 full power years, had an early history of fuel clad failures and was known as a high source term plant. Dose rates in the loop areas in Containment were approximately 1000 to 2000 mrem/hr with surface contamination levels averaging in the 10,000 to 100,000 dpm/100 cm2 range.

Routinely-accessed areas of the PAB, Spray and Fuel Buildings had dose rates of 10 to 50 mrem/hr, walkways were kept less than 1000 dpm/100 cm2, and equipment spaces had dose rates of up to 1000 mrem/hr and contamination levels on average of 5000 to 50,000 dpm/100 cm2. The LSA, RCA Storage and LLWS Buildings had dose rates of 10 to 200 mrem/hr depending on the type and quantity of waste in storage and contamination levels ranged from 5000 to 50,000 dpm/100 cm2 in liquid waste processing areas to less than 1000 dpm/100 cm2 in walkways.

Normal system leakage was responsible for the contamination levels found within the Containment, Spray, Fuel and Primary Auxiliary Buildings. Secondary plant areas were kept uncontaminated with the exception of a few components (e.g., component cooling system filters and steam generator blowdown demineralizer) which gave general area dose rates of a few mrem/hr. Primary and secondary component cooling systems were known to contain small amounts of residual Cs-137 from minor heat exchanger leakage which occurred during power operations. The auxiliary boilers and auxiliary condensate receiver also showed evidence of minor contamination from heat exchanger leakage which occurred early in the plants operating history.

In the late 1980s and early 1990s the plant began measures to reduce both the source term and surface contamination levels. Floor to ceiling area decontaminations were undertaken. High efficiency filters were installed in primary systems. One primary system chemical decontamination was performed which reduced primary system piping radioactivity levels by a factor of two.

MYAPC License Termination Plan Page 2-7 Revision 3 October 15, 2002 In 1990, the plant experienced a primary to secondary steam generator tube leak.

Prompt operator actions limited the secondary plant contamination. Following the steam generator tube leak, secondary systems were extensively surveyed during recovery activities and no residual activity was identified. Temporary controlled areas were established in the turbine hall to work on RCP motors, and the turbine hall sumps have indicated detectable plant nuclides.

The plant was shutdown in December 1996 for evaluation of cable separation problems. During the extended outage, economic conditions led to the decision to permanently shutdown in August 1997. A second chemical decon was performed following the decision to decommission the plant. The decontamination factors for the second decon improved to five to ten which resulted in loop area dose rates in the range of 50 to 200 mrem/hr. Contamination levels throughout the plant remained consistent with pre-shutdown values.

2.2.5 Current Radiological Status All fuel has been removed from the reactor and placed in the spent fuel pool, or l transferred to the ISFSI. The fuel pool has been converted to alternate cooling l and other primary systems have been drained and vented for decommissioning.

Chemical and Volume Control System waste resins and filters have been removed for disposal. The reactor vessel contained approximately 33,660 gallons of l slightly contaminated water. An additional 320,000 gallons added to the refueling l cavity for shielding during reactor component removal, have been processed as l radwaste.

MY does not expect any primary systems to remain after decommissioning. It is l expected that the diffuser will remain in place. Characterization of the diffuser is l described in Section 2.5.3 and Attachment 2H. Demolition of structures to 3 feet l below grade will remove the majority of embedded or buried piping. Remaining embedded or buried piping will be classified and surveyed in accordance with l Sections 2 and 5. l Based on both the Historical Site Assessment and the characterization surveys performed, a large portion of the site located to the West of Bailey Cove and North of the Ferry Road was determined to be non-impacted in the partial site l release applications (Maine Yankee Letters dated August 16, 2001 (MN-01-034) l and November 19, 2001 (MN-01-044) Early Release of Backlands (Combined) l Proposed Change 211, Supplements 1 and 2 respectively). The NRC granted the l request license amendment in its letter to Maine Yankee, dated July 30, 2002. l (See Attachment 2A and References, Section 2.7.) l Containment and control measures have prevented the release of radioactive material beyond the Bailey Point area as evidenced by no detection of plant-

MYAPC License Termination Plan Page 2-8 Revision 3 October 15, 2002 derived radionuclides above background levels in any of the measurements taken in or on the land area West of Bailey Cove and North of the Ferry Road. The same control measures will remain in effect during the decommissioning to prevent migration of contamination into clean or non-impacted areas.

The impacted areas of the site extend from the Ferry Road in a southerly direction down Bailey Point.

2.2.6 Hazardous and Chemical Material Contamination During its operational lifetime, MY used chemicals typical of steam power-generating facilities. In September 1998, MY had only non-bulk quantities of chemical and solvent waste stored on site awaiting disposal and no mixed wastes were in storage.

Preparation for decommissioning of the plant included removal of hazardous and chemical materials from plant systems. In 1998, 16,000 gallons of sodium hydroxide solution were removed from the spray chemical addition tank (SCAT) and neutralized, and chromates were removed from the water in the neutron shield tank using a totally-enclosed ion exchange resin process. A majority of the asbestos insulation was removed as part of the asbestos abatement project completed in January of 1999. Maintenance chemicals and hazardous materials were removed as specific plant areas were prepared for dismantlement.

Decommissioning of the plant includes removal of additional known l contaminants in plant systems and structures. Mercury switches, lead components, and PCB light ballasts are some examples of hazardous materials that are removed along with other plant components. Polychlorinated biphenyls l (PCBs) found at other nuclear facilities are also present at MY but are limited to painted surfaces and in some cable insulation material. Asbestos abatement continues to play a part in the removal of various components and building l materials. Section 3.6 of this LTP describes the coordination of activities with other agencies with regard to these contaminants.

Over the operational lifetime of the plant, spills to the environment occurred and l were generally cleaned immediately. In 1988, the facility experienced a 12,000 gallon chromated water leak from an underground component cooling pipe. Following repair of the leak, monitoring wells were installed and the extent of contamination and the effectiveness of remediation were monitored to the satisfaction of the Maine Department of Environmental Protection (MDEP). In 1991, one of the main transformers shorted and released approximately 200 gallons of transformer oil to the Back River. The spill was remediated to MDEPs satisfaction following the event.

MYAPC License Termination Plan Page 2-9 Revision 3 October 15, 2002 In these areas and throughout the site, MY will continue to work with the EPA and MDEP to demonstrate that areas have been adequately characterized, remediated if necessary, and are sufficiently clean to insure public health and safety. The EPA is supporting the Maine Yankee decommissioning project in several areas. The EPA is enabled by the Resource Conservation and Recovery Act (RCRA) to administer closure of facilities that were hazardous waste generators. Since the State of Maine Department of Environmental Protection has been delegated authority to administer the RCRA program in Maine, EPA is serving in a technical support role for the Maine Yankee site closure. EPA is expected to review all major closure-related documents and advise MDEP on their adequacy.

The EPA also is responsible for the Toxic Substances Control Act (TSCA), which serves as the primary means by which the use and disposal of PCBs and PCB-containing materials are controlled. PCBs have been identified above the TSCA limits of 50 parts per million (ppm) in electrical cable sheathing and, in limited areas, paint.

The MDEP has been delegated authority, by the EPA, to administer the National Pollutant Discharge Elimination System (NPDES) permit program as authorized by the Clean Water Act. Maine Yankee maintained an NPDES permit during operation.

2.3 Site Characterization Survey Methods As discussed in Section 2.1, the sites initial characterization survey work (ICS) was l performed by GTS Duratek and its subcontractor. Continuing Characterization Surveys l (CCS) were, and continue to be, performed by Maine Yankee, initially supported by the l former Decommissioning Operations Contractor (DOC), Stone & Webster (SWEC), and l its subcontractor, Radiological Services, Inc. (RSI). The FSS plan was based on this l information. These site characterization efforts used similar, but not identical, methods l and techniques. These differences are noted within the methods and results sections of this report.

2.3.1 Organization and Responsibilities GTS Duratek (GTS) was the prime contractor for the initial characterization surveys conducted from the fall of 1997 through the spring of 1998. GTS supplied hand- held instrumentation and performed field surveys. Subcontractors provided the following specialized services.

C IT Corporation performed the hazardous materials characterization survey and drive-over scans.

MYAPC License Termination Plan Page 2-10 Revision 3 October 15, 2002 C Duke Engineering & Services performed the activation analysis.

C Canberra Industries provided on-site laboratory instruments.

C Team Associates performed the asbestos characterization.

C Quanterra performed off site laboratory analyses.

Continuing characterization (CCS) activities began in the fall of 1998 and will l continue through decommissioning. Samples were collected and on-site surveys and analyses performed. Laboratory analyses for the hard-to-detect radionuclides were performed by Duke Engineering Services.

2.3.2 Characterization Data Categories Survey categories for initial site characterization (ICS) were designated by GTS as l surfaces and structures, systems, and environs (soils, sub-slab soils, sediments and groundwater) for both affected and unaffected locations based on the likelihood of the area being contaminated. The same designations are used for clarity and ease of comparing data.

a. Surfaces and Structures This category included building interiors and exteriors with associated structures, and, where applicable, the exterior surfaces of plant systems and components because these surfaces have the same potential for residual levels of radioactive material as the building surfaces in which they are located. Surface and structure survey packages also included ancillary buildings and structures. Structural material background measurements were also included in this category. These measurements were intended to determine general background levels for various building materials. If background reference area measurements are required for final survey measurements, they will be performed in accordance with Section 5.0.

In total, the survey category included approximately 7,900 measurements l in unaffected areas and approximately 6,400 measurements in affected l areas. This intentional bias toward unaffected surfaces and structures ensured no unsurveyed or undetected locations were likely to exist.

Affected structure surveys included 18 concrete core samples. Because concrete basement surfaces represent the key remaining structures upon l license termination, an additional 51 concrete core samples were obtained to improve nuclide data. (See Section 2.5.3 and Attachments 2F and 2G l for additional detail on these concrete cores and results.) l

MYAPC License Termination Plan Page 2-11 Revision 3 October 15, 2002

b. Systems This category included interior surfaces of process piping, components, ventilation ductwork, and installed drains and sumps. The levels of radioactive material on the internal surfaces of plant systems and components primarily depend on process operations. Therefore, these survey packages were separate from surface and structure survey packages.

Plant system survey packages generally were limited to one plant system. l This survey category included approximately 3,800 unaffected system measurements and approximately 1,050 affected system measurements.

Again the surveys were biased toward the unaffected systems to provide a high likelihood of identifying any existing contaminated pipe or component.

Additional systems surveys were conducted in order to bound the extent of contaminated components within non-Restricted Area structures.

c. Environs Land areas were surveyed and sampled to detect the presence and extent of soil contamination. Approximately one-third of the 820-acre site (original 740 acres + buffer land purchased later) land area received a gamma scan.

Measurements taken over the entire property used a grid system to adequately locate survey points. Nearly 300 soil samples were taken, 180 of which were from unaffected areas. One survey package in this category was devoted to obtaining background soil and exposure measurements from an area similar in physical characteristics to, but located several miles from, the site.

A study was performed to determine the amount of radioactivity present in the vegetation above the soil surface. Comparison measurements of soil and overlying vegetation showed no radionuclide activity in the vegetation exceeding background levels. FSS soil samples are therefore taken with overlying vegetation removed but with the root ball intact in accordance with approved procedures.

Sediment, groundwater and surface water samples were also included in this category. Over 100 sediment samples were obtained from shorelines, outfalls, catch basins, runoff ditches and the forebay. Twelve sediment samples were also obtained from offsite sources such as the Damariscotta River and Harpswell for background purposes. Over fifteen water samples were taken from groundwater monitoring wells, sumps, catch basins and

MYAPC License Termination Plan Page 2-12 Revision 3 October 15, 2002 an outfall. Five water samples were taken from offsite or unaffected sources for background purposes. In addition, the Radiological Environmental Monitoring Program has collected over 27 years of sediment, groundwater and surface water sampling data. For instance, the Annual Radiological Environmental Operating Report for 1999, submitted to the NRC on April 27, 2000, describes the automatic composite sampler located at the discharge of the forebay to monitor water discharged to the Back River. Samples were collected at least every two hours and subsequently composited for analysis. Groundwater from an on-site location was monitored quarterly. Shoreline sediment cores were collected semiannually from two locations on Bailey Point.

Multiple soil samples were taken and composited to determine the amounts and ratios of the hard-to-detect radionuclides in the most contaminated soils onsite.

Scan and fixed surveys of pavement were performed to identify potential sub-surface contamination. Two areas of soil contamination beneath pavement were documented in the HSA. One area of sub-slab leakage from the liquid waste effluent line occurred underneath the Service Building floor. The results of this soil contamination were contained in the 50.75(g) file.

2.3.3 Characterization Survey Design All phases of the characterization surveys were designed to sample each structure, l system and land area onsite for the presence of radioactive contamination. A heavy emphasis was placed on non-affected (non-impacted) systems, structures and areas with 2750 more surveys taken on non-affected systems, 1500 more surveys taken on non-affected surfaces and structures, and 18 survey packages devoted to non-affected areas versus 7 for affected areas. This emphasis ensured that the full nature and extent of the contamination were identified and characterized.

The initial radiological characterization survey (ICS) was organized, performed l and reported in one of five Groups and 127 packages which are listed in Section 2.3.7. Each group is comprised of plant areas containing similar types of media, or material, and similar contamination potential. The types of media included surfaces, structures, systems and environs. The environs category included facility grounds within and outside the RA, the liquid effluent pathway, Montsweag Bay, groundwater wells and remote locations within the MY Atomic Power Plant site boundaries. The contamination potential for the media in a given group was generally categorized as affected and unaffected. Affected areas had medium to high potential for containing contamination. Unaffected areas had a

MYAPC License Termination Plan Page 2-13 Revision 3 October 15, 2002 low or no potential for containing contamination. The affected/unaffected designation was not intended to indicate final survey classification status, but was intended as a general descriptor of contamination potential. The methods for converting any of the characterization survey results to classification of plant areas l for final site survey are described in Section 5 of this LTP.

Each group was further subdivided into survey packages that correspond to specific plant areas with similar operational history or physical location. The survey package breakdown is contained in Attachment 2B. All plant areas are l included in one of the survey groups/packages. The five groups are listed below.

C Group A-Affected Surfaces and Structures

• Group B-Unaffected Surfaces and Structures

• Group C-Affected Systems

• Group D-Unaffected Systems

• Group R-Radiologically Affected or Unaffected Environs These group designators were also used during continued characterization (CCS) l for survey package identification. Non-radiological data were collected and grouped into one of the following two categories listed below. The environs hazardous material characterization surveys (ICS) included testing for PCBs, l RCRA metals, semi- volatile organic compounds and volatile organic compounds.

C Group E-Hazardous Materials on Structures, Systems or Surfaces C Group H-Hazardous Materials in Environs Activation analysis calculations were also performed for the reactor vessel, reactor internals and the shield wall surrounding the reactor.

2.3.4 Instrumentation and Minimum Detectable Concentrations (MDCs)

Instrument Selection and Use Instrument selection, use and calibration for the MY characterization surveys (ICS l and CCS) were based on the assumed radionuclide mix and were performed in l accordance with approved procedures. Instruments used and their MDCs are described in the applicable section.

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a. Survey Methods Direct measurements of structures were performed with 126 cm2 gas flow proportional detectors for beta contamination. The MDC was between 500-2000 dpm/100 cm2 (as compared to the screening values of 5,000-11,000 dpm/100 cm2 ). The detector was kept within 1 cm of the surface.

Measurements of surface activity on small or restricted access areas were made using small Geiger-Mueller detectors or an array of multiple detectors for large bore systems or components. Measurement times were controlled in order to achieve the required MDCs.

Scan surveys were performed on both surfaces and land in order to detect areas of elevated activity for further investigation.

GTS Duratek performed scans (ICS) of open land areas with a 1 inch by 1 l inch NaI detector or the large drive-around plastic scintillator. Scan speeds were controlled in order to meet the required MDCs. Audible output was used with the handheld instruments to aid the surveyor in identifying areas of elevated readings. Continuing characterization scans (CCS) were performed using a 2 inch by 2 inch detector swept in a l pendulum pattern at a distance of 2 inches from the surface at a rate of 0.5 m/sec.

Samples of building materials, sediments, sludges and water were taken and analyzed using standard procedures and laboratory instruments. Smears for removable contamination were taken using standard techniques and laboratory counters. Exposure rates at one meter were measured using a NaI detector and a pressurized ion chamber. Soil samples of approximately 1000 g were cleaned to remove large debris and dried to remove moisture.

Samples were counted in Maranelli beakers using GeLi detectors for gamma emitters. Samples were analyzed by off site labs for Hard-To-Detect (HTD) radionuclides.

b. Minimum Detectable Concentrations for Volumetric Measurements The MDCs listed in Table 2-2 were typical values for both initial characterization (ICS) and continued characterization (CCS) samples, l which included HTD nuclides. The lower values were for gamma spec analyses. When characterization soil samples (ICS and CCS) were l analyzed for HTDs, the MDCs were maintained at levels as low as practicable.

Minimum detectable concentrations (MDCs) were defined for measurements and analyses used to quantify soil and other volumetric activity. Similar instruments, procedures, and MDCs applied to

MYAPC License Termination Plan Page 2-15 Revision 3 October 15, 2002 continuing characterization. MDCs for volumetric soil were less than 0.01 pCi/g for gamma nuclides versus a screening value of approximately 3-4 l pCi/g for a 10 mrem/yr annual dose. MDAs for Volumetric Water were l less than 2,500 pCi/L for H-3. There is no water screening value.

Table 2-2 Volumetric MDCs Type of Analysis MDC (pCi/g)

GTS DOC/MY l (ICS) (CCS)

Gamma Spectroscopy 0.10 0.01 - 0.1 Liquid Scintillation 2.0 to 3.0 2.5 Alpha Spectroscopy 0.10 0.01 to1.0 Radio Chemical

• 1 - 20 pCi/g

• 1 - 20 pCi/g Analysis

• except Ni-59

c. Structure and Surface Scan Sensitivities GTS Duratek used a slightly different method for calculating scan sensitivities (ICS) than the method specified in NUREG-1575/NUREG- l 1507. This approach increased the calculated scan MDCs by a factor of approximately 2.4. The use of this alternate approach had no effect on the interpretation and use of initial characterization data (ICS) . The l technicians evaluated detectably elevated readings during scan surveys based on changes in count rates regardless of the estimated MDC.

GTS Duratek performed a computerized sort of the direct measurements of total beta activity obtained during the characterization survey (ICS) of l unaffected areas by detector type, efficiency, local area background and use (building surfaces vs. system internals) in order to evaluate scan MDCs. The surface scan MDCs ranged from 2100 dpm/100 cm2 for large area gas flow detectors to 16,000 dpm/100 cm2 for system internals surveys.

The NUREG-1575/NUREG-1507 method was used to calculate scan sensitivities in the continuing characterization work (CCS) . This method l yielded surface scan MDCs of 1200-16,000 dpm/100 cm2 depending on the instrument and material being surveyed.

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d. Open Land Area Scans GTS technicians performed gamma scans of open land areas (ICS) using a l Ludlum 44-2, 1 inch by 1 inch NaI detector, and a TSA Systems Limited large area plastic scintillator, VRM-1X. (See Table 2-3.) In accessible areas, the VRM-1X detector, a 1.5 inch thick, by 3 inch wide, by 33 inch long block of scintillator-impregnated plastic, was the detector of choice because it had the lower theoretical MDC. The relatively large surface area of the VRM-1X detector greatly improves the probability of detecting isolated areas that contain elevated levels of radioactive materials.

Table 2-3 Theoretical Scanning Sensitivities Minimum Detectable Instrument Concentration/Activity Ludlum 44-2 14 pCi/g (Cs-137 source)

VRM-1X 11 pCi/g* (Distributed Co-60)

SPA-3 5 pCi/g (Cs-137 source)

• MDC as determined by Dr. Chabot in a letter to P. Dostie dated 11/12/98 Although GTS did not perform a priori MDC calculations, theoretical minimum detectable concentrations or minimum detectable activities for scans (ICS) performed with a vehicle-mounted VRM-1X detector, l traveling at less than 5 mph, were calculated for several geometries based on empirical data and numerical integrations following land surveys.

These data were examined by Dr. Chabot on 11/12/98 and found to be accurate within a factor of 2 to 4.

The SPA-3 detectors (2 inch by 2 inch NaI) were used for land area scans during continuing characterization (CCS) with scan MDCs of l approximately 5 pCi/g. This nominal MDC value of 5 pCi/g was based on l a background of 10,000 c/m, an index of sensitivity (d) of 1.38, a l surveyor efficiency factor of 0.707, and a conversion factor of 1200 c/m l per microR/hr, as stated in the manufacturers literature. The exposure l rate of soil for 5 pCi/g was determined by Microshield and was the same l value of 1.3 microR/hr, as given in Section 6.7.2.1 of NUREG-1575. l

MYAPC License Termination Plan Page 2-17 Revision 3 October 15, 2002

e. Instrument Calibrations Analytical and field instruments for both ICS and CCS were calibrated l using National Institute of Standards and Technology traceable sources representative of the assumed radionuclide mix at the MY site.

Instruments were calibrated at the MY site and, for GTS, at the GTS Duratek Central Calibration Facility in Oak Ridge, Tennessee or by vendors in accordance with the GTS Duratek Quality Assurance Project Plan for Site Characterization (ICS). Approved procedures were employed l to specify on-site instrumentation calibration requirements for continuing characterization (CCS). The average energy of the beta particles in the l MY radionuclide mixture was calculated. Based on the calculated average source beta energy of 0.088 Mev, Tc-99 (ave. beta energy of 0.085 Mev) was chosen for calibration. All of the alpha emitters have similar energies and Am-241 was chosen for the alpha calibration source. Tc-99 and Am-241 sources were used for calibrating gas flow proportional instruments used to perform surface scans and direct measurements. Cs-137 sources were used to calibrate exposure rate and soil scan instruments. The calibration program ensured that equipment was of the proper type, range, accuracy and precision to provide data to support the MY site characterization activities. The response of exposure rate and soil scan instruments to Co-60 was also determined during continued characterization (CCS) in order to detect discrete Co-60 particles. l 2.3.5 Quality Assurance Quality Assurance plans were developed for characterization work (ICS and l CCS). The elements of these plans were very similar. Differences between plans l are discussed below.

The GTS Quality Assurance Project Plan (QAPP) described the quality assurance requirements for the initial site characterization survey (ICS) . The QAPP l included applicable criteria from the GTS Duratek Quality Management System Manual specific to the MY project. The plan addressed sample collection, field survey measurements, sample analysis, data analysis/verification, and document control.

Continuing characterization (CCS) was performed using an approved CCS Quality l Control procedure which addressed the quality elements for these surveys. The procedure covered the requirements and frequency for replicate measurements, sample recounts, split samples, instrument use and control, sample custody, data verification/control, document control and investigation of unusual results.

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a. Quality Control Samples and Measurements For each laboratory instrument used during both initial characterization l (ICS) and continuing characterization (CCS) , laboratory personnel kept l daily quality control charts, a log of samples analyzed to provide traceability for each step of the analysis, and a maintenance log. Daily quality control checks were compared to specified tolerances. Control charts were developed at the time of initial calibration using a statistical analysis of repetitive measurements. Laboratory personnel maintained control charts for energy, full width at half maximum (FWHM), and efficiency for each gamma spectroscopy system and performed trend analysis daily. Routine background and blank counts demonstrated that the detector or cave had not become contaminated and confirmed sample detection levels. Daily checks were also performed on the analytical balance which was used to weigh the samples. Instruments failing the daily checks were removed from service until repaired.

The GTS Sample Analysis and Data Management Plan (ICS) identified l required quality control samples and measurements. In addition to the daily instrument quality control described above, laboratory personnel used quality control samples and measurements to verify system performance and data reproducibility.

The following on site QC analyses were performed and compared by GTS (ICS) using criteria in US NRC Inspection Procedure 84750: l C 10% of all samples were analyzed twice in the on-site laboratory (duplicate analysis)

C 10% of all samples were split and analyzed as two separate samples Quality control at the contract (off site) laboratories (ICS) also included l daily instrument checks and quality control samples that were analyzed during analysis of a batch of samples. Quality control samples and analyses for a batch of 20 (or fewer) samples analyzed by the contract laboratory included: a blank sample, a matrix spike sample (laboratory control sample, LCS), and a homogenized split sample. Laboratory control samples and analyses performed by the off-site laboratory were required to meet a relative percent difference (RPD) of 20% in accordance with the laboratorys internal procedures.

An approved CCS Quality Control procedure for the sample quality control criteria was developed. This procedure covered instrument daily checks, split or spiked sample requirements and acceptability criteria. Five percent of all survey units were chosen for repeat surveys with 10% of scans and

MYAPC License Termination Plan Page 2-19 Revision 3 October 15, 2002 fixed point measurements being replicated. Agreement for replicates was considered to be values within + 2 standard deviations. Instruments not passing the daily source check requirements were tagged Do Not Use and were removed from service until repaired. Data not meeting the replicate count criteria were removed from the data base until evaluated by an FSS specialist or engineer.

Duke Engineering & Services Environmental Laboratory performed laboratory analyses (CCS) under the requirements of DESEL Manual 100, l Laboratory Quality Assurance Plan.

The methods used by the off site laboratory for analysis of hazardous materials (ICS) were based on the EPA method for solid waste analysis l SW-846. Specific quality control samples, analysis, and acceptance criteria are specified in the analysis methods.

GTS personnel implemented the QAPP (ICS) through: l

• Scheduled audits and surveillances by on-site and off-site personnel

• Development of training matrices and training of personnel

• Development of records flow schedules

• Development of document control criteria

• Completion of readiness review checklists Self-assessments for CCS were implemented in accordance with approved Radiation Protection Performance Assessment Program procedures.

Training and qualification of survey personnel were assessed in accordance with the approved procedure for Selection, Training and Qualification of Radiation Protection Personnel. Records Control was maintained in accordance with approved procedures for QA Records Management.

b. Audits and Surveillances MY provided oversight of survey and sample activities to determine whether the overall characterization plan was implemented as designed. l External audits of project activities included assessments by MY personnel and subcontractors. These included an audit of the GTS Duratek facility (ICS) in Kingston, TN and project-specific audits based on the Quality l Assurance Program Plan and other project plans. These audits did not identify any project-specific nonconformances. In addition, MY personnel

MYAPC License Termination Plan Page 2-20 Revision 3 October 15, 2002 and their contractors performed surveillances on daily project operations.

Characterization personnel identified, tracked, and corrected concerns generated by these surveillances.

MY Radiological Engineering and GTS Duratek corporate and Field Services personnel (ICS) performed internal audits of the project. Also, at l the request of MY, GTS Duratek appointed an on-site surveillance technician. This inspector, trained on quality assurance procedures, performed daily surveillances on project activities. Characterization personnel (ICS) tracked and corrected nonconformances identified by these l surveillances according to approved procedures.

During continued characterization (CCS), audits and self assessments were l performed on the characterization activities. The results of the findings were entered into the trend data base and tracked to resolution in accordance with the approved procedure for the Corrective Action Program.

2.3.6 Data Quality Objectives Initial site characterization (ICS) was planned prior to the issuance of NUREG- l 1575. However, a retrospective look at site characterization revealed that Data Quality Objectives (DQOs) 1, 2, 3 and 4 were addressed by GTS Duratek. The characterization plan identified the problem, the decision method, the resources, the team, the decision makers, the sample requirements, the instrumentation and MDCs, the expected nuclides, the survey areas and basic data analysis. While the use of a formal DQO process may have resulted in a more efficient characterization process, the resulting data have been shown to be sufficient to meet the objectives listed in Section 1.0 and are therefore acceptable.

The DQO process was used during continuing characterization (CCS) to meet the l objectives outlined in Section 2.1. Contamination boundaries, radionuclide profiles, data standard deviations and projected sample sizes were determined during continuing characterization.

Data Quality Objectives 5, 6 and 7 are addressed in LTP Section 5, Final Status Survey, and Section 6, Compliance with the Radiological Criteria. In particular for DQO 5, the parameter of interest is specified as the mean of the residual contamination level in a survey unit, the action levels include the DCGL and the investigation levels, and the decision rule is described for the determination to release a survey unit. For DQO 6, the limitations of decision errors are addressed by specifying the respective probabilities of making a Type I and Type II decision error, the lower boundary of the grey region (LBGR) and the minimum value for relative shift. For DQO 7, the survey design for collecting data is optimized by

MYAPC License Termination Plan Page 2-21 Revision 3 October 15, 2002 using exposure pathway modeling to develop some site-specific DCGLs, adjusting the LBGR to obtain the optimum relative shift, evaluating survey instrumentation and measurement techniques and selecting appropriate actions following the exceedance of investigation levels.

2.3.7 Survey Findings And Results The results of the initial characterization surveys (ICS) are reported by survey l group and package number as identified below.3 Site and Survey Area maps are provided in this section of the LTP to graphically depict the boundaries of each area. These maps are not drawn to scale but are sufficient to show the presence of areas of high contamination.

3 Additional survey packages were developed (and are discussed in this section) as necessary to support data collection for continued characterization. These later packages are not listed here in Section 2.3.7.

MYAPC License Termination Plan Page 2-22 Revision 3 October 15, 2002 PACKAGE GROUP A Affected Structures and Surfaces Survey Packages NUMBER A0100 Containment Building - Elevation -2 ft.

A0200 Containment Building - Elevation -20 ft.

A0300 Containment Building - Elevation 46 ft A0400 Fuel Building - Elevation 21 ft.

A0500 Demineralized Water Storage Tank TK Elevation 21 ft.

A0600 Primary Auxiliary Building - Elevation 11 ft.

A0700 Primary Auxiliary Building - Elevation 21 ft.

A0800 Primary Auxiliary Building - Elevation 36 ft.

A0900 Service Building Hot Side - Elevation 21 ft.

A1100 Low Level Waste Storage Building - Elevation 21 ft.

A1200 RCA Building - Elevation 21 ft.

A1300 Equipment Hatch Area - Elevation 21 ft.

A1400 Personnel Hatch Area - Elevation 21 ft.

A1500 Mechanical Penetration Room - Elevation 21 ft.

A1600 Electrical Penetration Room - All Elevations A1700 Containment Spray Building - All Elevations A1800 Auxiliary Feed Pump Room - Elevation 21 ft.

A1900 HV-9 Area - Elevation 21 ft.

A2100 Refueling Water Storage Tank (RWST) TK Elevation 21 ft.

A2200 Borated Water Storage Tank (BWST) - Elevation 21 ft.

A2300 Processed (Primary)Water Storage Tank (PWST) - Elevation 21 ft.

A2400 Test Tanks 14A/14B -Elevation 21 ft.

A9900 Concrete core contamination profile sampling A9901 Activation analysis core sampling A9902 Activation analysis core sampling

MYAPC License Termination Plan Page 2-23 Revision 3 October 15, 2002 PACKAGE GROUP B Unaffected Structures and Surfaces Survey Packages NUMBER B0100 Turbine Deck - Elevation 61 ft.

B0200 Old Control Room - Elevation 21 ft.

B0300 Motor Control Center (MCC)/Battery Room - Elevation 62 ft.

B0400 Fire Pump House - Elevation 1 B0500 Condenser Bay - Elevation 21 ft.

B0600 Condenser Bay - Elevation 39 ft.

B0700 Service Building Cold Side - Elevation 21 ft.

B0800 Fuel Oil Building - Elevation 21 ft.

B0900 Emergency Diesel Generators - Elevation 21 ft.

B1000 Auxiliary Boiler Room - Elevation 21 ft.

B1100 Recirculating Water Pump House - All Elevations B1200 Administration Center - Elevation 21 ft.

B1300 WART Building - All Elevations B1400 Visitor and Information Center - Elevation 1 B1500 Warehouse 2 - Elevation 1 B1600 Training Annex Building - Elevation 1 B1700 Staff Building - All Elevations B1800 Spare Generator Building - Elevation 1 B1900 Environmental Services Building - All Elevations B2000 Bailey Barn - Elevation 1 B2100 Lube Oil Storage Room - Turbine Building Elevation 21 ft.

B2200 Cold Machine Shop - Turbine Building Elevation 21 ft.

B2300 Cable Vault Room - Turbine Building Elevation 39 ft.

B2400 Staff Building Tunnel - Staff Building to Turbine Building Elevation 21 ft.

B9800 Structural Background Survey

MYAPC License Termination Plan Page 2-24 Revision 3 October 15, 2002 PACKAGE GROUP C Affected Plant Systems Survey Packages NUMBER C0100 Primary and Post Accident Sampling System C0200 Waste Solidification System C0300 Containment Spray System C0400 Emergency Core Cooling System C0500 Residual Heat Removal System C0600 Primary Vents and Drains C0700 Fuel Pool Cooling System C0800 Waste Gas Disposal System C0900 Pressurizer and Pressurizer Relief System C1100 Reactor Coolant System C1200 Boron Recovery System C1300 Chemical and Volume Control System C1400 Liquid Waste Disposal System C1500 Primary Auxiliary Building Drains C1600 Primary Auxiliary Building Ventilation C1800 Containment Ventilation System C1900 Steam Generators

MYAPC License Termination Plan Page 2-25 Revision 3 October 15, 2002 PACKAGE GROUP D Unaffected Plant Systems Survey Packages NUMBER D0100 Condensate System D0200 Water Treatment Plant Systems D0300 Potable Water System D0400 Sanitary Sewer System D0500 Circulating Water and Screen Wash System D0600 Service Water System D0700 Fire Protection System D0800 Lube Oil System D0900 Compressed Air System D1000 Auxiliary Boiler System D1100 Steam Generator System D1200 Main and Reheat Steam System D1300 Auxiliary Steam System D1400 Main Turbine and Turbine Control System D1500 Steam Dump and Turbine Bypass System D1600 Main Feedwater System D1700 Emergency/Auxiliary Feedwater System D1800 Heater Drain and Extraction Steam System D1900 Component Cooling Water System D2000 Vacuum Priming and Air Removal System D2100 Amertap System D2200 Secondary Plant Sealing System D2300 Auxiliary Diesel Generator D2400 Secondary Sample and Chemical Addition System D2500 High Pressure Drains D2600 Environmental Services Laboratory Systems D2700 Administration Building HVAC System D2800 Information Building HVAC System D2900 Turbine Building Ventilation System D3000 Staff Building HVAC System D3100 Service Building HVAC System D3200 Hydrogen and Nitrogen System D3300 Turbine Building Sumps and Drains

MYAPC License Termination Plan Page 2-26 Revision 3 October 15, 2002 PACKAGE GROUP D Unaffected Plant Systems Survey Packages NUMBER D3400 Low Level Radioactive Waste Storage Facility

MYAPC License Termination Plan Page 2-27 Revision 3 October 15, 2002 PACKAGE GROUP R Environs Affected and Unaffected Survey Packages NUMBER AFFECTED R0100 RCA portion (West Side) of Protected Area Yard R0200 Balance of Protected Area (East Side)

R0300 Roof and Yard Drains #006, #007 and #008 R0400 Forebay Area Shorelines R0500 Bailey Point R0600 Ball Field R0700 Construction Debris Landfill UNAFFECTED R0800 Administration and Parking Areas R0900 Balance of Plant Areas R1000 Foxbird Island R1100 Roof and Yard Drains #005, #009-12, #017 and N-12 R1200 Low Level Radioactive Waste (LLRW) Storage Building Yard R1300 Dry Cask Storage Area R1400 Westport, Montsweag Bay, Bailey Point Cove and Plant Area Shorelines R1500 Ash Road Area Rubble Piles R1600 Owner Controlled Area West of Bailey Cove R1700 Owner Controlled Area North of Old Ferry Road R1800 Bailey House Area R1900 Bailey Cove R2000 Diffusers R2100 Maintenance Yard (Stockyard)

R2200 Background R2300 SFPI Substation Slab R2400 IT Duplicate Samples R2500 Driveover Elevated Areas R2501 Follow-up sampling at Elevated Soil Sample Locations (south of Refueling Water Storage Tank and Contractor Parking Lot)

R2800 10 CFR 61 Analysis Sampling Hazardous and chemical material surveys (ICS) were performed on the materials, l systems and areas as specified in the tables for Group E and Group H below. The

MYAPC License Termination Plan Page 2-28 Revision 3 October 15, 2002 data for these groups are presented in the Summary of Site Characterization Data section which follows.

PACKAGE GROUP E Plant Surfaces, Structures and Systems Hazardous Material NUMBER Survey Packages E0100 Protected Area Paint E0200 Plant Electric Components E0300 Transformer Oils E0400 Plant Pump Oils E0500 Various Plant Fluids E0600 Component Cooling Water E0700 Brass, Bronze and Cadmium Plated Components E0800 Plant Batteries E0900 Mercury Components E1000 Asbestos Insulation and Other Materials E1100 Asbestos Containing Components E1200 Lead Shielding E1300 Paint Outside Protected Area

MYAPC License Termination Plan Page 2-29 Revision 3 October 15, 2002 PACKAGE GROUP H Environs Areas Hazardous Material Survey Packages NUMBER H0100 Oil and Hazardous Material Transfer and Handling Areas (4)

H0200 Diesel Oil Tank Loading Area H0300 Main, North, Spare and Shutdown Transformers H0400 Roof and Yard Drains #006, #007 and #008 H0500 Solid Waste Storage Area H0600 Primary and Secondary Side Waste Storage Building Yard Areas H0700 Drumming/Decontamination Waste Accumulation Area H0800 Diffuser Forebay H0900 Reactor Water Storage Tank Area H1000 Groundwater Monitoring Wells B-201 through 206, MW-100, BK-1 H1100 Warehouse Yards H1200 Fire Pond and Yard Area H1300 Construction Debris Landfill H1400 Bailey Point H1500 Administration and Parking Areas H1600 Roof and Yard Drains #005, #009-12 and N-12 H1700 Surface Flow Drain #005 H1800 Balance of Plant Area H1900 Foxbird Island H2000 Low Level Waste Storage Yard H2100 Dry Cask Area H2200 Environmental Services Laboratory H2300 Switchyards H2400 Areas Outside Plant Impact

MYAPC License Termination Plan Page 2-30 Revision 3 October 15, 2002 2.4 Summary of Initial Characterization Survey (ICS) Results l The operational history and the range of contamination determined during initial site l characterization (ICS) are summarized in this section for the survey groups indicated l above. More detailed data including mean, maximum, and standard deviation are presented by survey package in Attachment 2B. l 2.4.1 Group A Affected Structures and Surfaces Group A includes buildings and surfaces within the RA including levels of the Reactor Containment, Fuel, and Primary Auxiliary Buildings, as well as tanks containing radioactive liquids, electrical/mechanical penetration areas and concrete surface samples. Areas of known contamination with very high dose rates were sampled less than areas with more moderate dose rates in order to maintain the exposure to surveyors ALARA. Survey data were taken from posted areas which included High Radiation Areas, Radiation Areas, Radioactive Material Storage Areas and Contaminated Areas. These areas include the reactor coolant system and waste processing equipment and are among the most highly contaminated areas on site. However, several locations within this group contained no radioactive systems, components and structures or were found to be below station limits for posting as contaminated (viz., DWST, PWST, electrical and mechanical penetration areas and the auxiliary feed pump room).

Maximum total surface activities ranged from greater than 100,000 dpm/100 cm2 in the RCA Building, Containment Building (CTMT), and Spray Buildings to less than 1000 dpm/100 cm2 in auxiliary support areas (e.g., electrical/mechanical penetrations). Maximum removable beta activities ranged from greater than 128,000 dpm/100 cm2 in the CTMT to less than MDA in auxiliary support areas.

No removable alpha sample activities were above the MDA values which indicated little or no transuranic (TRU) surface contamination. Maximum net exposure rates reported in Attachment 2B ranged from about 4,000 µR/hr in the Primary l Auxiliary Building (PAB) to around 5 µR/hr in the mechanical penetration area.

Operational surveys reported containment exposure rates ranging from 1 mrem/hr to over 1000 mrem/hr.

Group A results combined with the operational survey data and knowledge of process provided the information needed to target those structures within the RA requiring remediation, establish radionuclide profiles and provide estimated radioactive waste volumes.

MYAPC License Termination Plan Page 2-31 Revision 3 October 15, 2002 2.4.2 Group B Unaffected Structures and Surfaces Group B was comprised of buildings and surfaces located outside the RA including the Turbine Hall, sections of the Service Building, the Control Room, office spaces and various out buildings such as the Fire Pond Pump House, the warehouse, and the Bailey House/Barn. With the exception of a few closed secondary systems and a few locations in the Turbine Hall, Service Building and warehouse, none of these buildings contained or stored radioactive material during plant operation and are therefore some of the lowest activity areas on site. Sealed sources for instrument calibration were stored at the Bailey House environmental laboratory.

The crane bay and turbine deck in the Turbine Hall were used for RCP motor refurbishment. The 1990 steam generator tube leak affected steam and feedwater components in the Turbine Hall. The auxiliary boilers were known to be internally contaminated. Some areas within the Service Building such as the old decon shower and primary chemistry lab sample hoods were also known to be slightly contaminated. The warehouse was used as a shipment and receipt point for small quantities of packaged radioactive material. There was no evidence of leakage detected at the warehouse from packages shipped or received.

Maximum total surface activities ranged from a high values of 3700 dpm/100 cm2 and 8600 dpm/100 cm2 in the Turbine Building (certain floor areas) to lows of

<1000 dpm/100 cm2 in outlying areas, such as the cable vault. The Ball Field Dugout indicated 700 dpm/100 cm2, which was later identified by the State of Maine as Co-60. Maximum removable beta activities ranged from 200 dpm/100 cm2 in the Turbine Building to less than MDA in other areas. No areas had plant related alpha activity above the MDA level. Maximum exposure rates ranged from 26 µR/hr in the Service Building to 2 µR/hr in the Turbine Building. Tritium was detected slightly above MDA in several water-containing systems. High beta readings in the Bailey House were confirmed to be NORM from the granite foundation blocks.

Group B surveys verified that most of the Turbine Hall was free of residual radioactivity. Continuing characterization surveys (CCS) established the extent l and limits of radioactivity in the areas in which it was found.

2.4.3 Group C Affected Plant Systems This group was comprised of the radioactive systems such as the RCS, CVCS, ECCS, liquid and solid waste, containment ventilation and primary vents and drains. The survey packages in this group consisted of systems and components that will be removed and disposed of as radioactive waste during decommissioning and, therefore, do not require characterization to support Final Status Survey (FSS). These are the highest radioactively contaminated systems at MY.

MYAPC License Termination Plan Page 2-32 Revision 3 October 15, 2002 Total surface activities were not measured on these systems internals, as their activity levels were too high. Instead, 15 cm and 1 meter external exposure rate measurements were taken at four quadrants from system locations, to support dose to curie calculations, for waste shipping purposes. Internal system surfaces of the steam generators were found to be contaminated up to 500,000 dpm/100 cm2 removable beta activity. Alpha activity was present at as much as 35 dpm/100 cm2 in the CVCS indicating possible TRU contamination. Exposure rates in these areas ranged from a low of 13 µR/hr in the Waste Solidification system to more than 16,000,000 µR/hr in the Spent Fuel Cooling and Refueling system.

Group C results verified the extent of contamination in primary systems and provided data needed to support the Radiation Protection Program during component removal in addition to providing information needed for waste classification.

2.4.4 Group D Unaffected Plant Systems Including the Sewage Treatment System This group consisted of secondary side systems that were designed to remain non-contaminated. Examples of these systems are main steam, feedwater, compressed air and potable water. However, certain parts of the secondary side systems contained minor levels of contamination. The auxiliary condensate system was l known to be slightly contaminated due to aux boiler problems early in plant life.

Turbine Hall sumps were known to be slightly contaminated due to reactor coolant pump motor refurbishment activities taking place in the Turbine Hall. Steam and feedwater systems were potentially impacted by the 1990 steam generator tube leak. The Service Water system was impacted by liquid effluents from the Test Tanks. Several of the systems crossed over to the RA, where elevated readings were detected in/on the systems but were later attributed to NORM interference in the analyses. Group D systems were generally the lowest in activity of all those surveyed.

Until the early 1980s when they were disconnected, hot side shower drains and toilets were directed to the sewage treatment plant. Initial characterization surveys l (ICS) showed elevated readings in one hotside shower drain. In the two years l following shutdown, routine chemistry analyses of both the on site holdup tank and l the municipal treatment facility have shown no plant-derived radionuclides.

Radionuclides have been detected in the sewage plant as a result of employees receiving medical isotope therapy.

Survey results from Group D established the limit and extent of residual activity in systems expected to be clean and provided information to properly control the systems as well as classify the waste during decommissioning. Some of the systems in Group D had elevated readings indicating the possible presence of plant

MYAPC License Termination Plan Page 2-33 Revision 3 October 15, 2002 derived radioactive material. Further measurements were made on these systems as part of the continuing characterization (CCS) plan to properly evaluate the level l and extent of contamination. These measurements support release and/or disposal l determinations.

2.4.5 Group R Environs Affected and Unaffected The group was broken down into 7 affected and 18 unaffected areas. Environs sampling covered all areas of the 820 acre site (740 acres original site + purchased buffer properties). Fifteen of the sample areas showed no detectable plant derived radioactivity. Ten of the areas (R0100, R0200, R0300, R0400, R1000, R2000 and R2300 within the protected area and R0500, R0900 and R1300 outside the protected area but on Bailey Point) had elevated readings requiring further evaluation and sampling.

Asphalt, sub-asphalt soil and uncovered soil to the South and West of Containment, Spray, Fuel and RCA Storage Buildings were known to be contaminated by system leaks and radioactive waste container storage. Excavated soil and asphalt from the RA were temporarily placed on Bailey Point and later returned to the RA. Silt from condenser cooling water intakes was removed and spread on site land located to the north and west of the 345 kV electrical switch yard. Plant-derived radionuclides had been detected in estuary sediments as a result of permitted liquid releases by environmental samples (REMP reports) taken at various times during plant operation. Minor contamination was located near storm drains adjacent to the RA. Contamination levels ranged from 1pCi/g to 11 pCi/g for Co-60 and 1pCi/g to 156 pCi/g for Cs-137 in the areas of known soil contamination from old leaks/spills (R0100).

Marine sediment samples were obtained from shorelines, outfalls of catch basins, runoff ditches and the forebay. In addition, the Radiological Environmental Monitoring Program had collected over 27 years of sediment sampling data.

Shoreline sediment cores were collected semiannually from two locations off Foxbird Island. Additional sampling of off-site marine sediments will be conducted pursuant to an agreement between Maine Yankee and Friends of the Coast (FERC Offer of Settlement dated December 31, 1998.)

Survey packages with indications of potentially elevated activity levels (R0500, R0600, R0700, R0800, R1000, R1300, R1600 and R1800) were combined into an investigation package designated R2500. The highest levels of activity were detected on Bailey Point from the investigation package R2500 (up to 34,000 pCi/g l of Co-60) and the activity was remediated during sampling. Follow up samples taken in three areas after remediation of detected activity were documented in package R2501.

MYAPC License Termination Plan Page 2-34 Revision 3 October 15, 2002 Three areas (R1500, R1600, R1700) were classified as non-impacted based on operational data, the Historical Site Assessment and the initial characterization l (ICS) results. l Group R surveys determined which land areas were non-impacted and which were impacted. This group also provided the information necessary to project waste volumes from contaminated soils.

2.4.6 Ventilation Ducts and Drains Results for the biased sampling of building vents and drains can be found within the survey data for Groups C, D and R. Ventilation ducts and system drains were sampled as the most likely collection point for system contamination. This biased sampling provided a high level of assurance that contaminated systems were located, identified and, when found within secondary side buildings, marked to provide the necessary level of control over radioactive material.

Affected System Vents and Drains (C0600, C1500, C1600 and C1800) showed mean removable contamination values ranging from 53 to 51,000 dpm/100 cm2 and maximum values from 6000 to 140,000 dpm/100 cm2.

Unaffected System Vents and Drains (D1800, D2000, D2500, D2700, D2800, D2900, D3000, D3100 and D3300) had two systems positively identify residual radioactivity. The Service Building HVAC (D3100) had significant activity above the MDA which was due to the hot side ventilation sources going to the Service Building ventilation duct work. D3000 Turbine Building Sumps and Drains had two (2) sumps test positive for plant derived nuclides (up to 1.7pCi/g Co-60). The Sump Oil Collection Tanks (TK-91) also test positive (1.1 pCi/g Co-60). There were four (4) other systems (D1800 - Heater Drain Extraction Steam, D2700 -

Admin Building HVAC, D2900- Turbine Building Ventilation, and D3000 - Staff Building HVAC) with elevated activity. However, the elevated readings were likely due to radon daughter activity. This will be confirmed during CCS and/or the operational free release program. The High Pressure Drains showed tritium l activity at levels just above MDA. Tritium in these areas have been attributed to NORM interference in the analyses.

Survey results from this group established the limit and extent of residual radioactivity in systems and provided necessary information for properly controlling material and for proper classification of waste during decommissioning.

2.4.7 Buried and Embedded Piping A review of prints and drawings was performed during CCS to determine the amount of buried and embedded pipe. MY has a limited amount of piping actually

MYAPC License Termination Plan Page 2-35 Revision 3 October 15, 2002 embedded in concrete. Total embedded piping includes approximately 800 feet of primary and secondary component cooling water pipes. Based on inventory l estimates made in 2002, the total embedded piping expected to remain on site is l approximately 940 linear feet, representing slightly over 150 m2. A detailed listing l of the embedded piping inventory is provided in Attachment 6-7. l Component cooling piping showed maximum activity up to 22,000 dpm/100 cm2 and will be removed during demolition activities. Small segments of refueling cavity and spent fuel pool skimmer piping (approximately 175 feet) are embedded within the walls of the two pools. The skimmer piping is known to be contaminated and activity levels could be as high as 20,000 to 180,000 dpm/100 cm2 removable beta contamination based on data obtained from spent fuel pool cooling (C0700) and RHR (C0500) survey packages. This piping will be removed.

Circulating water and service water pipes are buried cast concrete pipes rather than embedded pipes. Eighteen direct measurements above MDC were identified in the circulating water pipes. Service water discharge piping receives the liquid effluent overboard pipe with approximately a 3 foot embedded section and showed maximum activity levels of 3100 dpm dpm/100 cm2 of removable beta contamination. Mean values were less than MDA.

Embedded piping above the 17 foot elevation will be removed. Pipes below 17 feet will either be removed during demolition or will be properly evaluated to ensure compliance with the enhanced state standards of 10 mrem/yr for all pathways including not more than 4 mrem/yr from groundwater sources of drinking water. Maine Yankee has produced an informational set of site drawings showing the as left condition after decommissioning. These drawings identify the remaining buried or embedded pipe, conduit, building penetrations, cable vaults, and duct banks. This set of drawings will be used to plan FSS surveys.

The following describes the principal sections of buried and embedded piping l which is expected to remain following decommissioning and which will be decontaminated as necessary and subject to FSS.

a. Containment Spray Piping and CS Valves-approximately 68 ft. (C0300): l During plant operation, the system was filled with reactor coolant water.

Initial site characterization surveys (ICS) identified this as a contaminated l system. Gamma isotopic samples collected from the system identified the presence of plant-derived nuclides (Co-60 and Cs-137). The portion of the system that will remain following demolition of above grade structures is embedded in the concrete foundation of the Containment Building. Two valves from the containment spray system are also encased in concrete.

Levels up to 40,000 dpm/100 cm2 were detected in the spray system (C0300) during ICS. Higher levels of contamination have been found in l

MYAPC License Termination Plan Page 2-36 Revision 3 October 15, 2002 subsequent surveys. This 16 inch embedded piping makes up a surface area l of 26.5 m2. l

b. Containment Foundation Drains-approximately 378 feet.(C2000)4: The l foundation drain system was used to transfer groundwater from around the Containment Building foundation to lower the hydrostatic pressure on the foundation. The system consists of four partially embedded transfer pipes that drain to the foundation sump. The system has a high potential for residual contamination. The drain system is wholly contained within the RA and has been subjected to liquid spills in the soil around the Containment Building. The system was not surveyed during initial site l characterization (ICS); however, the sump water was sampled periodically. l Tritium is the only nuclide identified in the sump water at levels exceeding natural background. A water sample was submitted for HTD analysis during CCS and only tritium was detected. See Section 2.4.12. No l removable surface contamination or direct surface measurements have been made. This combination of 2 inch and 6 inch embedded piping makes a l surface area of 30.2 m2. l

c. Sanitary Waste (D0400): A portion of the sanitary waste piping is buried l beneath the Turbine Hall floor slab and extends to the sewage treatment plant. At one time early in the plants operation, the pipe transferred waste from sanitary facilities located within the RA. The original discharge point for treated sanitary waste was into the circulating water inlet bay. In the mid-1980s, the sanitary system was connected to the town of Wiscasset sewage system. The sanitary system, including the discharge to the town of Wiscasset, has been sampled periodically since the plant began operation.

Radionuclides detected in recent years were limited to medical isotopes which are short lived and would not be present by the time the system pipe is surveyed. Of 37 fixed point surface measurements of the system taken during ICS, two were in the RA, and both indicated elevated activity of up to 5700 dpm/100 cm2. Both of these samples were from a disused drain in the system that will be removed during dismantlement. No removable contamination was identified in the system. Gamma isotopic samples from the system did not indicate the presence of plant-derived radionuclides.

d. Circulating Water System-approximately 1600 feet (D0500): The l circulating water system consists of 4 buried concrete inlet pipes which carried sea water from the Back River to the condenser then overboard to 4

As noted in Section 2.3.7, additional survey packages were developed for data collection during continued characterization (i.e., not part of ICS) and, thus, are not listed in Section 2.3.7. Survey Packages C2000, D3500, and D3700 are examples of packages developed for CCS and/or for FSS using the same numbering system as was used for ICS.

MYAPC License Termination Plan Page 2-37 Revision 3 October 15, 2002 the forebay and is finally discharged through a diffuser in the Back River, down stream of the inlet. The circulating water system is considered a secondary side system in that there was a physical barrier (condenser tubes and steam generator tubes) between the circulating water and the contaminated primary plant (reactor coolant system). The circulating water system has a very low potential for residual contamination. The operational history of the facility indicates no significant primary to secondary leakage occurred. Additionally, the circulating water system pressure was maintained above the pressure of the turbine exhaust steam in the condenser so that even if there was a condenser tube leak, it would have carried sea water into the condensate system. During Initial Site Characterization, low levels of detectable activity were identified on the main condenser outlet side of the circulating water system. The suspected cause of the contamination was recirculation of allowable effluent discharges into the suction side of the Circulating Water Pump House. The maximum fixed point total surface contamination measurement collected during ICS was 811 dpm/100 cm2. No removable contamination was identified in the system. Gamma isotopic samples collected in the system during ICS did not identify any plant-derived nuclides.

e. Service Water System (D0600): The Service Water System consists of two l buried inlet pipes which carried sea water through the component cooling heat exchangers. The discharge of the system consists of a single buried line which goes into the seal pit.

The discharge side of the pipe receives the liquid effluent discharge pipe.

During initial site characterization (ICS) , low levels of detectable activity l were identified on the discharge side of the piping. No direct beta measurements were above the MDA. Nine samples of removable beta activity were detected above the MDA (3134 dpm/100cm2 was the maximum value). The positive indications of residual activity in this system are associated with the liquid effluent header location and the liquid radwaste radiation monitor installed at that location. Gamma isotopic samples collected at the liquid effluent line entrance point and at the radiation monitor were positive for Co-60 (700 pCi/g). The waste header is contained within its own local Restricted Area within the Turbine Building.

The radwaste piping will be removed and disposed of as radioactive waste.

The remaining portions of the service water discharge piping meet the criteria of a Class 3 area.

f. Fire Protection (D0700): The water-filled portion of the fire protection l system is the only section that will remain following demolition. Water for firefighting was stored in a man-made storage pond located on site.

MYAPC License Termination Plan Page 2-38 Revision 3 October 15, 2002 Makeup water for the pond came from Montsweag Brook. (The storage pond is addressed as part of survey area R0900). The fire protection system was not piped to containment. The system consists of a loop of buried pipe which circles the yard and supplies various hydrants and headers. The fire protection system is considered a support system in that it did not interface with other operating systems (e.g., primary coolant or steam supply). The fire protection system has a very low potential for residual contamination. Although sections of the system did reside within the RA, system pressures were sufficient to prevent inleakage. The fire water system has been cross-connected with potentially contaminated systems in the past. However, samples collected during CCS have only identified naturally occurring radioactive material. The maximum fixed point total surface contamination measurement taken during ICS was 1116 dpm/100 cm2. Gamma isotopic samples collected during ICS did not identify any plant-derived radionuclides in the system.

g. Storm Drains (D3500): The Storm Drain (SD) system is used to drain storm l water and runoff from the facility to the Back River and Bailey Cove. The system functions as a gravity drain system to remove the water via a system of drain grates, manholes and system piping. The system drains the entire site both inside and outside the Protected Area. Manholes 1 through 3 (Section 1 of the system) drain the Protected Area outside the Restricted Area and south of the Turbine Building and Service Building. The outfall for this portion of the system is a 24 line that drains to the Back River south of the Circulating Water Pump House (CWPH). Manholes 4 and 5 (Section 2 of the system) drain an area inside the Protected Area outside the Restricted Area east of the Turbine Building. This line drains the area around the Main Transformers. The outfall for this leg of the system is a 15 line that drains to the Back River north of the CWPH. Manholes 6 through 11 and un-numbered manholes north of the Turbine Building (Section 3 of the system) drain an area both inside and outside the Protected Area. The area drained is all outside the Restricted Area. These legs all collect at Manhole 7 and the combined outfall is routed to the Back River immediately adjacent to the north side of the CWPH. Manholes 13 and 14 (Section 4 of the system) drain the upper access road and the upper contractor parking lot. The outfall for this section of the system is the Back River north of the Information Center building. Manholes 30A, and 31 through 37 (Section 5 of the system) drain an area inside the Protected Area in the Restricted Area. This leg of the system drains the main RCA Yard area around the Containment Building and the alley between the Containment Building and the Service Building. These legs all collect at Manhole 35 and the combined outfall is routed to the Forebay Seal Pit.

Manholes 21 through 24 (Section 6 of the system) drain the north side of the Restricted Area and the roof of the WART Building. The area drained

MYAPC License Termination Plan Page 2-39 Revision 3 October 15, 2002 is inside the Protected Area and both inside and outside the Restricted Area.

The combined outfall for this leg joins another leg at Manhole 27.

Manholes 25A, 25B, 26 through 29 and 38 (Section 7 of the system) drains areas adjoining the Fire Pond and Warehouse and outside the west end of the Restricted Area. The outfall from Manhole 24 joins this leg at Manhole

27. The combined outfall for this leg of the system is routed to Bailey Cove.

Samples collected during ICS and knowledge of process indicate that the Storm Drain system has a low potential in some legs and a high potential in some legs for residual contamination. Sections 1 through 4 have a low potential for residual contamination. Sections 5 through 7 have a high potential for residual contamination. Sections 1 through 4 drain areas that have historically been outside the Restricted Area and have a low potential for residual contamination. Sections 5 through 7 drain areas in and adjacent to the Restricted Area and may have become contaminated due to loose surface contamination in and on yard structures and equipment being washed into the drain legs by rain water runoff and snow melting.

Since the roof drains flow to the storm drains and the portions of the roof drains above 17 feet will be removed, the roof drains will be included in the storm drain survey.

h. Containment Building Penetrations (D3700) (411ft): Several Containment l Building penetrations will remain following demolition of the above grade structure. The penetrations contain embedded piping from numerous primary and secondary systems. The remaining penetrations are as follows:

- Approximately 20 linear feet of up to 1 piping l

- Approximately 35 linear feet of 1.5 piping l

- Approximately 50 linear feet of 2 piping l

- Approximately 35 linear feet of 3 piping l

- Approximately 55 linear feet of 4 piping l

- Approximately 100 linear feet of 6 piping l

- Approximately 45 linear feet of 8 piping l

- Approximately 5 linear feet of 10 piping l

- Approximately 25 linear feet of 16 piping l

- Approximately 10 linear feet of 24 piping l

- Approximately 20 linear feet of 30 piping

- Approximately 11 linear feet of 40 Fuel Transfer Tube piping l Each of these penetration, except for the Fuel Transfer Tube, consists of a l five foot length of pipe penetration through the containment foundation l wall. The calculated surface area of this embedded piping is approximately l 78 m2. l

MYAPC License Termination Plan Page 2-40 Revision 3 October 15, 2002

i. The Primary Auxiliary Building and Spray Building Penetrations (60ft). l Several non-containment piping penetrations through the Primary Auxiliary l Building and Spray Building will remain in the respective building l foundations following demolition of the above grade structure. Each of l these penetrations consists of a 2 to 3 foot length of pipe penetration l through the building foundation wall. The calculated surface area of this l embedded piping is approximately 19.5 m2. l

j. The spent fuel pool liner leak detection system (24ft). Four 1 inch lines l embedded in the spent fuel pool structure will remain following demolition l of the above grade structure. The calculated surface area of this embedded l piping is approximately 1 m2. l The penetrations that will remain in the Containment Building have a high potential for residual contamination. One of the systems identified as having a remaining section of embedded piping is Containment Spray, which is known to contain residual contamination.

ICS data collected in the Containment Spray system (C0300) indicate the presence of removable contamination and gamma isotopic samples identified the presence of plant related radionuclides. ICS were not collected in the Fuel Transfer Tube. Additionally, no specific contamination controls have been established for the remaining sections of the embedded piping and the majority of the Containment Building is posted and controlled as a surface contamination area.

2.4.8 Asphalt, Gravel and Concrete Two site locations containing asphalt and gravel from non-RA construction work were sampled for activity (R0700 and R1500). Neither location showed activity above background for plant-derived nuclides.

Because of the potential impact of concrete on the exposure pathway, concrete core samples were collected and analyzed during initial characterization (ICS) l (A9900, A9901, A9902) and continuing characterization (CCS) . In 1998, GTS l Duratek took seven (7) concrete core samples that were later subjected to analysis l by Stone and Webster to determine HTD nuclides at low MDCs. In 1999, forty- l three (43) additional concrete core samples were obtained and analyzed by gamma l spectrometry. In 2000, an additional eight (8) concrete cores were collected and l analyzed for HTD nuclides at low MDCs. Table 2C-2 lists the original 43 cores l (1-1A through 11-2A) taken during continuing characterization plus the 8 l additional cores (12-1A through 13-3A) collected in 2000 for a total of 51 cores. l Three of the cores (3-1A through 3-3A) were activated concrete and are labeled as l

MYAPC License Termination Plan Page 2-41 Revision 3 October 15, 2002 activation samples in Table 2C-2. Four samples (5-6A, 6-5A, 6-6A, and 7-2A) l had no reported activity. Section 2.5.3a discusses the establishment of the nuclide l mixture for contaminated concrete surfaces. See Attachment 2F for a description l of the process used to evaluate the concrete surface nuclide mixture. See l Attachment 2G for additional discussion of concrete core sample collection and l processing. l Concrete activity was found to be due to penetration of surface contamination as well as activation of concrete constituents in areas exposed to neutron flux.

(Activated concrete comprised approximately 5% of the concrete in containment.) l Surface contamination penetration was primarily limited to the top 0.1 cm.

Activation activity generally followed expected activation curves, peaking at 1 to 2 inches into the concrete, and dropping off at greater depths (A9902). Slight anomalies in concrete activation were noted in the vicinity of embedded rebar.

Positive indications of activation were seen as deep as 24 inches in some concrete samples that were exposed to high neutron fluence. As noted in Section 3.3.3, l activated concrete will be removed down to the activated concrete DCGL.

As part of CCS, samples of local fill material (sand, gravel, and till) were analyzed for bulk density and Kd. Activated Concrete at levels above the activated concrete DCGL will be removed.

2.4.9 Paved Areas One paved area near the warehouse (R0900) exhibited one elevated exposure reading. A small contaminated area was removed during sample collection and was found to contain a small amount of Co-60. Resurvey confirmed removal of the contamination. Paved areas within the RA are known to have sub surface asphalt and sub surface soil contamination as described in the Historical Site Assessment section.

2.4.10 Components The status of individual components is given in the systems data, Groups C and D.

Group C components are found in radioactive systems and are known to be contaminated.

Section 2.4.3 describes the affected components in Group C; Section 2.4.4 describes the unaffected components in Group D, and Attachment 2B provides a l detailed summary of components during ICS.

MYAPC License Termination Plan Page 2-42 Revision 3 October 15, 2002 2.4.11 Structures, Systems and Environs Surveyed For Hazardous Material (Groups E and H)

These surveys identified expected amounts of waste chemicals, lubricants and solvents; toxic metals in switches; and PCBs in paints and cables. Some areas of soil contamination by motor oils/fuels were discovered which will require further evaluation. Initial characterization activities (ICS) confirmed the presence of lead- l based paint and PCBs in both cables and paints. Several small areas of soil were found to be contaminated by chemical or hazardous material.

Hazardous material health and safety considerations will be assessed through the RCRA closure process described in Section 8.6.2. l 2.4.12 Surface and Groundwater ICS sample results for surface and groundwater were reported within the individual survey area packages (R0100, R0200, R0300, R1100, R2200 and R2400) and are summarized in Attachment 2B. l Tritium was the only plant derived radionuclide detected in groundwater and surface water during ICS. The overall range of the tritium analyses was <793 pCi/L to 6812 pCi/L. The highest value was from the Containment foundation sump. All of the measurements were well below the EPA Drinking Water MCL of 20,000 pCi/L. The Containment foundation sump is currently being monitored and trended as part of CCS to determine if there is evidence of plant derived tritium contamination in the groundwater.

2.4.13 Background ICS measurements were made of several types of construction materials from offsite locations which were used as background samples. Soil samples from remote locations were also taken and analyzed to be used as background soils.

ICS material backgrounds (concrete, brick, ceramic, etc.) were subtracted from reported ICS data direct measurements of total beta activity. ICS environs background (soil, sediment, water, etc.) were collected for informational purposes only. ICS environs background data were not subtracted from ICS environs survey reported data.

a. Material Background The natural levels of radioactivity in plant construction materials affected direct measurements for total beta activity. To quantify this effect, GTS Duratek performed a background study (ICS) at the Central Maine Power l

MYAPC License Termination Plan Page 2-43 Revision 3 October 15, 2002 Headquarters Building in Augusta, Maine. The study included direct measurements for total beta activity on painted and unpainted concrete and concrete block, ceramic tile, and asphalt. Other materials encountered during the initial characterization survey (ICS) such as glass, carpeting, and l steel were not included in the background study since their natural radioactivity would not contribute significantly to direct measurements for total beta activity. Survey personnel used the same instruments for the structural background survey as were used for the initial characterization l survey (ICS) . Count times were adjusted to ensure minimum detectable l activities of approximately 300 dpm/100 cm2. Project personnel used these results to correct data gathered from similar surfaces during the initial l characterization survey (ICS) . l The following is a summary of ICS material backgrounds:

Table 2-4 Summary of ICS Material Backgrounds MATERIAL AVERAGE (dpm/100cm2)

Bare Concrete (& block) 665 Painted Concrete (& block) 478 Asphalt 925 Ceramic Tile 1109 Other (duct, bare & painted metal, etc.) 0

b. Environs Background The purpose of the environs background study was to measure and document the levels of radionuclides, especially Cs-137, present in local soils and typical background exposure rates. The survey sampling and measurement techniques complied with approved procedures and supporting guidance documentation. Sample materials for the background study included surface soils, sediments and groundwater. The project team performed gamma spectroscopy for all samples, and analyzed groundwater for tritium. The average Cs-137 concentration in soils was determined from samples collected at the Merrymeeting Airfield, from a hay field, woodlands, and scrub lands. The average Cs-137 concentration in marine sediments was determined from samples collected from the Damariscotta River, near Dodge Point and Harpswell. Groundwater concentrations were determined from the Eaton Barn, Bailey House, and Days Ferry. No

MYAPC License Termination Plan Page 2-44 Revision 3 October 15, 2002 groundwater samples had detectable Cs-137 or tritium concentrations (above MDA).

The survey also included an in situ gamma spectrum with a MicroSpec multichannel analyzer/sodium iodide detector. Survey technicians measured background exposure rates with a sodium iodide detector.

Additionally, the survey team took both sodium iodide and pressurized ion chamber (PIC) measurements at each of the background soil sample locations in the hay field at Merrymeeting Airfield to observe the energy response of the PIC versus the sodium iodide detector. The project team calculated the background exposure rate and PIC measurement ratio for information and did not use the results to adjust any other measurements.

The following is a summary of ICS environs background data:

Table 2-5 Summary of ICS Environs Background Data MEDIA MINIMUM MAXIMUM AVERAGE Sediment Cs-137 0.04 pCi/g 0.11 pCi/g 0.07 pCi/g Soil Cs-137 (Combined) 0.09 pCi/g 1.42 pCi/g 0.45 pCi/g Soil Cs-137 (Woodland) 0.1 pCi/g 0.92 pCi/g 0.52 pCi/g Soil Cs-137 (Hay Field) 0.1 pCi/g 0.55 pCi/g 0.38 pCi/g Soil Cs-137 (Scrub 0.09 pCi/g 1.42 pCi/g 0.55 pCi/g Lands)

Water H-3 <743 pCi/L <3126 pCi/L <2024 pCi/L Wood & Scrub Land 5.9 µR/hr 8.3 µR/hr 7.2 µR/hr l Exposure (NaI2) l Open Land Exposure 10.0 µR/hr 13.6 µR/hr 11.6 µR/hr l (NaI2)

Open Land Exposure 7.18 µR/hr 9.34 µR/hr 8.22 µR/hr (PIC)

c. Miscellaneous Background Survey Data The University of Maine (Dr. C. T. Hess) performed a radiological soil and sediment background study prior to plant operations and reported the data in EPA Technical Note ORP/EAD-76-3. The study included analysis of

MYAPC License Termination Plan Page 2-45 Revision 3 October 15, 2002 nine soil samples, two marine sediment samples, and seven water samples collected in the vicinity of Maine Yankee prior to plant operations in during 1972.

The following is a summary of miscellaneous background survey data:

Table 2-6 Summary of Miscellaneous Background Survey Data MEDIA MINIMUM MAXIMUM AVERAGE Sediment Cs-137 0.35 pCi/g 0.45 pCi/g 0.4 pCi/g Soil Cs-137 0.8 pCi/g 4.96 pCi/g 2.04 pCi/g Water H-3 <90 pCi/L <400 pCi/L <294 pCi/L l 2.4.14 Waste Volumes and Activities Table 3-8 summarizes projected activities associated with various sources of radioactive waste materials generated during decommissioning.

2.5 Continuing Characterization (CCS) l The sites initial characterization work (ICS) left a few survey areas unresolved with l respect to the nuclides present and the extent or boundaries of contamination. Those areas were characterized during the Continuing Characterization Survey (CCS) effort, which l included obtaining the following data: l

• Soil samples from the southeast fence area for bounding the extent of l contamination

• Soil samples from the contractors parking lot to confirm remediation and support construction of the ISFSI

• Soil samples from Bailey Point to confirm remediation

• PCC/SCC survey to bound the extent of contamination

• Condensate/Auxiliary Condensate survey to bound the extent of contamination

• Service Water survey to bound the extent of contamination

• Concrete cores l

MYAPC License Termination Plan Page 2-46 Revision 3 October 15, 2002 l

• Forebay/diffuser media l l

• Groundwater l l

The new Spent Fuel Pool Decay Heat Removal System is contaminated. Remediation plans call for the system components to be removed and disposed of as radwaste. Once fuel has been transferred to the ISFSI, the area occupied by the SFP cooling system will be surveyed. Additional sampling of the circulating water discharge Forebay was performed to assure compliance with specific unrestricted use release criteria.

As noted in Section 2.1, characterization samples (CCS) will continue to be collected and l analyzed throughout the project to support the need for the most current and accurate radionuclide data.

2.5.1 Methods Methods employed for continuing characterization were consistent with those described in Section 2.3 for site characterization. Any differences between the methods used by GTS (ICS) and the methods employed for Continuing l Characterization (CCS) are noted within Section 2.3. l The work was performed under the guidance of a Decommissioning Work Order (DWO) and in accordance with approved procedures. In order to ensure comparable results, the instrumentation used during CCS was similar in design, l function and sensitivity to that used during initial characterization.

2.5.2 Results The range of residual radioactivity existing on surfaces and within soils and systems targeted for sampling during Continuing Characterization (CCS) are l summarized below. Detailed data including mean, maximum, and standard deviation are presented by survey package in Attachment 2D. The standard l deviations calculated from CCS data may be replaced with more appropriate values calculated from post remediation or post demolition survey data. This section provides summary results from CCS. The current, resulting nuclide fractions are describe in Section 2.5.3.

a. Stone & Webster Review of the GTS Report (ICS) l Upon review of the GTS Duratech report (ICS), Stone & Webster identified l areas requiring additional characterization as follows: l

MYAPC License Termination Plan Page 2-47 Revision 3 October 15, 2002

1. Determine the extent of soil contamination at the Southwest fence (CR0200, CR10005) - The East/West boundaries of the soil contamination were determined by gamma spectroscopy of soil samples. In addition, soil was sent for radiochemical analyses in order to confirm the ratio of radionuclides including the hard-to-detect nuclides.

2. Verify remediation of the contractor parking lot contaminated areas (CR1300) - Contrary to the GTS report and prior to continued characterization activities commencing, the State of Maine reported that the soil in the parking lot still contained Co-60 contamination after remediation. Soil survey results verified that there was residual soil contamination. The contaminated soil was excavated and disposed of as radwaste. A sample matrix was developed for post-remediation surveys and soil samples were taken and counted. Following this cleanup, the parking lot was determined to be successfully remediated based on gamma spectroscopy of soil samples and gamma scans taken over the affected soil area.

3. Verify remediation of the Bailey Point soil storage area (CR0500) - A sample matrix was developed and soil samples were taken and counted. Based on gamma spectroscopy results, the Bailey Point soil storage area was determined to have been successfully remediated, pending final status survey.

4. Bound the extent of contamination in the PCC and SCC systems (CD1900) - PCC was opened and system internals were analyzed by gamma spectroscopy to determined the extent of contamination. The PCC system was found to be contaminated throughout, including the lube oil coolers of the diesel generators. The SCC system contamination was limited to one air conditioner feeding the control room (which had previously been in the PCC system but was later changed to SCC for train separation concerns) and both SCC 5

Note: Survey package numbers, as initially established for characterization, are listed in Section 2.3.7. l To distinguish a given packages data from the characterization phase to the Final Status Survey (FSS) l phase, a convention was adopted. A preceding C was added (to the package number) to indicate the l characterization and a preceding F would be used to denote the FSS phase of the project. Thus, l CR0200" in the LTP text refers to the survey package containing characterization data for survey l package R0200. l

MYAPC License Termination Plan Page 2-48 Revision 3 October 15, 2002 pump suction elbows. The systems were labeled to show the extent of contamination.

5. Bound the extent of contamination in the Condensate/Aux Condensate systems (CD0100) - Samples were taken from the aux condensate piping, aux condensate receiver, and aux boilers. The samples confirmed that the aux condensate piping and aux boilers were contaminated. The system was labeled to show the extent of contamination.

6. Bound the extent of contamination in the liquid waste discharge line as it enters the Service Water pipe (CD0600) -

Samples of the service water system were taken up stream from the point of entry of the liquid waste discharge pipe.

The samples confirmed that contamination was limited to the area adjacent to the discharge pipe connection.

7. Additional surveys were designed and implemented to resolve reported positive count rate data on various systems or components in the Turbine Hall.

The activity in the water treatment plant (CD0200) was determined to be l Naturally Occurring Radioactive Materials (NORM).

The data obtained during the Continued Characterization Surveys (CCS) are presented in Attachment 2C tables.

Data obtained during all phases of characterization surveys are used to l determine the nuclide profile for each media or material. If conditions arise during decommissioning which might affect the nuclide profile, additional sampling will be performed to verify the nuclide profile of any affected medium.

b. Soils Surface soil was sampled and analyzed for radionuclides during the initial site characterization (ICS). The radionuclides were detected in the top 15 l cm of on-site soil in the survey areas encompassing the backyard.

Additional data were collected during continued characterization to better establish nuclide profiles. The predominant plant-related, beta-gamma emitting radionuclides detected were H-3, Co-60, Ni-63 and Cs-137. Two sets of higher activity soil samples taken by GTS were composited and subjected to radiochemical analyses for the hard-to-detect nuclides. No TRUs were detected in the composites when analyzed with techniques

MYAPC License Termination Plan Page 2-49 Revision 3 October 15, 2002 giving MDAs of 0.01 pCi/g to 1.0 pCi/g. The actual soil nuclide profile is l provided in Section 2.5.3.

The samples from each area were analyzed by gamma spec. If the gamma spec results were consistent with reported values, between 240 and 800 g were removed from the sample containers and added to the composite. The amount removed depended on the total number of samples available from each location. The composites were well mixed and counted again to ensure expected results were achieved. The composites were then sent for HTD analysis except for H-3. Tritium was not analyzed because the samples had been in storage for a long time and were exceptionally dry.

Samples for H-3 analysis were taken from locations adjacent to the original sample locations. K-40 and Th were not reported because they were not plant-derived nuclides.

During characterization (CCS) a concern was raised about activity in the l vegetative layer of soil. As a result, a comparison was performed by counting vegetation and the soil/root ball; there was little measurable activity in the vegetation. Future soil samples will include the surface soil layer but not the protruding vegetation.

Sub-surface soil has been sampled and characterized in areas in which there was knowledge or indication of contamination below 15 cm. The nuclide ratios were consistent with surface ratios. In addition, building sub-slab soil characterization will be performed during remediation and demolition to determine the presence and extent of any sub-slab contamination.

Samples will be taken alongside foundation walls or through holes bored through the floor if necessary.

For additional discussion on soil samples and nuclide fraction see l Attachment 2I. l l

c. Systems and Components Residual contamination on or in plant piping was the result of the deposition of both fission and activation products. Prior to and during characterization surveys (both ICS and CCS), samples of process piping l were obtained to determine which systems were contaminated and the current radionuclide profiles including the hard-to-detect nuclides. The bounds of the contaminated piping were not established initially so systems were opened and surveyed to define the bounds of contamination.

Contaminated system components and piping will be removed and disposed of as radioactive waste.

MYAPC License Termination Plan Page 2-50 Revision 3 October 15, 2002 Fe-55, Ni-63, Co-60 and Cs-137 made up 99 percent of the system activities determined during initial characterization. TRUs contributed less than 1 percent of the total activity. The major beta-gamma emitter detected in system materials was Co-60 with a range of activity of 1 to 715 pCi/g (MDAs were 0.03 to 5 pCi/g). No additional quantitative gamma analyses for systems or components were conducted during CCS.

d. Buried and Embedded Piping Buried and embedded piping remaining after demolition will receive special surveys during the FSS. The nuclides and ratios in piping and contaminated components are consistent with those described in c above since the systems with embedded sections of contaminated pipe were the systems sampled during initial characterization. The nuclide profile is provided in Section 2.5.3. Nearly all of the embedded pipe consists of the l through-wall stubs of 1 to 4.5 feet in length. Since the embedded pipe l contributes approximately 2 tenths of one percent of the total annual dose l rate, it was decided to assume the small lengths of embedded pipe were l contaminated with the same source term as the concrete surfaces through l which they passed. Buried pipe is considered to be contaminated with the l same source term as other contaminated surfaces, and the activity is l released into the surrounding soil upon pipe degradation. Buried pipe l contributes less annual dose than embedded pipe. l

e. Structures-Concrete Concrete structures at elevations higher than 3 feet below grade will be demolished. Surfaces (at elevations below 3 feet below grade) will be decontaminated to the specified DCGL for unrestricted use criteria. (See Section 3 for details on building demolition.). Four radionuclides, Cs-137, Ni-63, Co-60 and H-3 comprise approximately 99 percent of the radioactivity on concrete surfaces. (Special consideration was given to trench and sump surfaces. See discussion in Section 2.5.3.)

Radioactivity found in the concrete shielding materials in containment was the result of both contamination and activation. Concrete cores were removed and analyzed in order to estimate the radioactivity levels and nuclide distributions of shielding materials. The predominant radionuclides present in structural (activated) concrete are H-3, Fe-55, Eu-152, C-14, and Co-60 (comprising approximately 98 percent of the activity in activated concrete).

Concrete cores were counted using both hand-held instruments and gamma spectrometers. This information, coupled with the radiochemical analytical

MYAPC License Termination Plan Page 2-51 Revision 3 October 15, 2002 data, were used to determine instrument total efficiency Et values (reported in Section 5.5.2).

f. Summary of CCS Activities Since Submittal of Revision 0 of the l LTP Since the submittal of Revision 0 of the LTP, several confirmatory samples have been collected. Two floor trench concrete samples were taken and submitted for HTD analysis to confirm or rule out some nuclide outliers reported by GTS (ICS) from a trench sample processed by another l laboratory.

Three additional Containment Building floor samples and three PAB floor samples were taken to replace the cores consumed during analysis. See l Attachment 2G for discussion of concrete core sample collection and l processing. l A portion of activated concrete with embedded rebar was sent for analysis on both the concrete and rebar to establish the hard-to-detect nuclide fraction. A comparison of the nuclide profile was made to activation analysis results prepared for MY activated material as well as to published activation data. The results compared favorably in both instances. A core from the in-core instrumentation (ICI) sump was extended to a depth of 22 inches in order to improve the activated concrete profile (i.e, variance of activity with depth; see Table 2-10). The depth profile will be used to plan remediation activities for the ICI sump area. The projected post-remediation activity remaining in the ICI sump area was used in the dose calculations described in Section 6.6.2.

Fire pond water samples were taken and analyzed for tritium and gamma emitters. The same was done for the reflecting pond and sediment from the pond was counted to well below environmental LLDs in order to show there were no plant-derived nuclides in the sediment. See Table 2C-3 for l results of reflecting pond samples. (Fire pond water and sediment results l are not included since the fire pond will be demolished.) l A containment foundation sump water sample was analyzed (including HTDs) to relatively low MDAs. Tritium was determined to be the sole nuclide present in the foundation drains and groundwater based on this analysis. (This finding was consistent with sump water monitoring results from the past years.) See Section 2.5.3.d for additional information l regarding site hydrogeology and groundwater sampling, and the l establishment of the groundwater nuclide fraction used for dose l assessment. l

MYAPC License Termination Plan Page 2-52 Revision 3 October 15, 2002 As part of both initial and continuing site characterization, forebay l sediment was sampled. To gain additional insight regarding the spatial l distribution of contamination and to support further characterization and l remediation planning, additional sampling efforts were undertaken. The l principal campaign was in Spring 2001 and included the sampling of: (1) l sediment around the protective rip-rap (inside the forebay), (2) underwater l sediment on the structure floors, (3) exposed material on the forebay ledges l near the weir wall, and (4) dike soil material beneath the rip-rap. Diver l operations and inspections of the diffuser also provided an opportunity for l the sampling of sediment inside the diffuser piping, as well as piping l coupons. The characterization of the forebay and diffuser system is l summarized in 2.5.3e and described in more detail in Attachment 2H. l Section 6.6.9 discusses the associated dose assessment related to these l contaminated media. l Additional material background samples were also collected in order to get better sample population statistics.

The results of these additional samples were used with previous data to determine nuclide profiles for each medium or material. In addition, detailed analyses of concrete core data were performed to ensure that the data collected were truly representative of the contaminated concrete on site. The soil and activated concrete data were also re-evaluated to confirm earlier assumptions based on the data reported in Revision 0 of the LTP.

2.5.3 Nuclide Profile One of the purposes of Site Characterization (both ICS and CCS) is to establish the l radionuclide profiles for the various contaminated media which provide dose to the critical group. Multiple samples were taken of each type of media in order to determine the nuclides present and their relative fractions to one another. These nuclide fractions are presented by media in the following sections.

a. Contaminated Concrete Surfaces (Including Special Areas) l Multiple concrete cores were analyzed (including HTDs) in order to determine the nuclide profile for contaminated concrete surfaces. The majority of the potentially contaminated surfaces remaining will be concrete. Other contaminated material, such as buried and embedded pipe, l may also remain. The nuclide profile determined for contaminated concrete is assumed to apply to all concrete surfaces. The sample results were l averaged over the entire population and the individual samples compared for consistency. As might be expected, the data were somewhat varied

MYAPC License Termination Plan Page 2-53 Revision 3 October 15, 2002 depending on the concrete location, spill history, decontamination history, surface coating and age.

The nuclide fraction for contaminated material was established using each of the positively identified nuclides. The non-detected nuclides were assumed not to be present in the mixture. In order to ensure that the elimination of non-detected nuclides at their MDC levels would not significantly affect the results, a sensitivity analysis based on dose was performed. Dose rates were determined for each individual core, for the core average values and for the average of the fractions using all nuclides in the suite at their actual value or their reported MDA, then the analysis was repeated using only the detected nuclides.

Two of the original set of nine cores (both containment floor trench l samples) showed evidence of TRUs; however, the values were very near the analytical MDCs. Even so, the TRUs were included in the evaluation of the nuclide fraction. Upon closer examination, the nuclide fraction for the trench samples appeared distinctly different from the other concrete fraction. The trench had a slightly different history of nuclide contact than the floor surfaces in general. Most significantly, water had been drained directly to the trench during the machining of cobalt-containing thermal shield pins and other special evolutions. Based on the sample results from l the two trench cores and consideration of the operational trench history, additional sample data were obtained to confirm the non-trench data. From that data, a separate nuclide fraction for the trenches was developed. As discussed Section 6.7, a separate DCGL for trenches was also established.

Additional concrete cores were taken and analyzed, revealing other areas in l the plant warranting a separate nuclide fraction. See discussion below l related to special areas. l Table 2-7 gives the nuclide fraction for contaminated surfaces that was selected based on the analysis of the characterization data determined by the average of the fractions method and decayed to 1/1/2004. Table 2-7 l provides the nuclide fraction for the balance of plant contaminated l concrete surfaces. l Table 2-8 gives the nuclide fraction for special areas in the plant. These l areas include the containment outer annulus trench, the PAB pipe tunnel, l and the letdown heat exchanger cubicle. These were separated from the l balance of plant contaminated concrete surfaces and were chosen based l on operating conditions and the presence of TRU contamination. The dose l consequences and DCGL for this collection of areas are described in l Section 6.7.2. l

MYAPC License Termination Plan Page 2-54 Revision 3 October 15, 2002 The data variability for the concrete cores was analyzed on the basis of l dose. The significance of any identified variability was judged on its effect l on the resulting dose. (See Attachment 2F for detailed discussion of the l data analysis.) l Table 2-7 Nuclide Fractions Contaminated Concrete Surfaces (Balance of Plant Areas) l Nuclide Fraction (as of 1/1/2004)

H-3 2.36E-2 Fe-55 4.81E-3 Co-57 3.06E-4 Co-60 5.84E-2 Ni-63 3.55E-1 Sr-90 2.80E-3 Cs-134 4.55E-3 Cs-137 5.50E-1

MYAPC License Termination Plan Page 2-55 Revision 3 October 15, 2002 Table 2-8 Nuclide Fractions for Contaminated Concrete Surfaces l Special Areas l Nuclide Nuclide Fraction (1/04)

Mn-54 4.03E-04 l Fe-55 2.24E-02 l Co-60 3.64E-01 l Ni-63 3.02E-01 l Sr-90 6.87E-03 l Sb-125 4.52E-03 l Cs-134 2.82E-03 l Cs-137 2.89E-01 l Pu-238 1.17E-04 l Pu-239 8.75E-05 l Pu-240 8.75E-05 l Pu-241 6.71E-03 l Am-241 5.93E-04 l Cm-243 4.65E-05 l Cm-244 4.45E-05 l

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b. Activated Concrete / Rebar Activated nuclide ratios were found to be consistent with published values.

The major variation with activated concrete was a decrease in total activity with depth in the material as shown by two deep core profile samples. This property can be used to determine the depth of remediation needed. There was also a local effect on nuclide activity and ratio in the area immediately surrounding rebar contained within the concrete.

Two highly activated concrete samples were analyzed for HTDs. As noted in Section 2.5.2f, one portion of activated concrete included embedded l rebar. The rebar sample was also analyzed for HTDs. The hard to detect nuclides showed the same level of consistency as the gamma emitters when compared to published values (NUREG/CR-3474). The nuclide fractions for the activated concrete and rebar was established using each of the positively identified nuclides. The non-detected nuclides were assumed not to be present in the mixture. In order to ensure that the elimination of non-detected nuclides at their MDC levels would not significantly affect the results, an analysis based on dose contribution was performed. Annual dose rates were determined for each nuclide at its actual reported value or its MDC, then the analysis was repeated using only the actual reported values of the detected nuclides. Those nuclides included in the dose analysis at their MDC values were shown to contribute less than 10 percent of the annual dose from the pathway analyzed. Table 2-9 gives the nuclide fraction for activated concrete and rebar decayed to 1/1/2004.

Based on the higher dose contributions from activated concrete, in comparison to the rebar, the nuclide fraction for activated concrete was used in the Section 6 dose assessment. See Section 6.6.2.

The activated concrete (and rebar) will be removed until a value of 1 pCi/g l is reached. This value will be determined by volumetric sampling (LTP l Section 5.5.1.a). If, in the course of this sampling, the results show a l different nuclide fraction, an evaluation will be performed to determine if l changes to the nuclide fraction are warranted. Dose impact will be a key l consideration in this evaluation. l

MYAPC License Termination Plan Page 2-57 Revision 3 October 15, 2002 Table 2-9 Activated Concrete Nuclide Fractions Concrete as of Rebar as of 1/2004 1/2004 Nuclide Fraction Fraction H-3 0.647 -------

C-14 0.058 -------

Fe-55 0.124 0.910 Ni-63 0.007 0.006 Co-60 0.040 0.084 Cs-134 0.0084 --------

Eu-152 0.111 --------

Eu-154 0.009 --------

Table 2-10 shows the activity measured a function of depth in the deep core sample.

MYAPC License Termination Plan Page 2-58 Revision 3 October 15, 2002 Table 2-10 Activated Concrete: Deep Core Sample Activity Profile Depth (in)*

• Activity (pCi/g)* *

• Depth (in) Activity (pCi/g) l 0 - 0.5 677* 10.75 - 11.5 87 l 0.5 - 1.0 828 11.5 - 12.25 23 1.0 - 1.5 845 12.25 - 13.0 23 1.5 - 4.0 824 13.0 - 13.75 17 l 4.0 - 4.75 771 13.75 - 14.5 14 4.75 - 5.5 329 14.5 - 15.25 14 5.5 - 6.25 534 15.25 - 16.0 11 6.25 - 7.0 365 16.0 - 16.75 7 7.0 - 7.75 290 16.75 - 17.5 6 7.75 - 8.5 233 17.5 - 18.25 6 8.5 - 9.25 206 18.25 - 19.0 1 9.25 - 10.0 182 19.0 - 20.0 1 10.0 - 10.75 103

• Adjusted to remove Cs-137 surface contamination from the total activity l

• • Note that the depth column represents a label for each sequential slice and is not intended as an exact measurement. The slices were generally 1/2" to 3/4" but were not uniform in thickness. Therefore, while Table 2-10 presents the profile out to 20 inches, this represents all of the data available for the entire 22 inch core.

• • • Measured activity provided in this table includes gamma detectable l activity from the nuclides listed in Table 2-9. l

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c. Contaminated Soil Soil from the areas with the highest contamination levels (RWST and PWST areas) were composited and analyzed for nuclide content including HTDs.6 Since the samples used for the composites were very dry, archived soils, no tritium analyses were made. However, tritium analyses were performed on soil samples from an adjacent area.

The nuclide fraction for the contaminated soil was established using each of the positively identified nuclides. The non-detected nuclides were assumed not to be present in the mixture. In order to ensure that the elimination of non-detected nuclides at their MDC levels would not significantly affect the results, an analysis based on dose contribution was performed. Annual dose rates were determined for each nuclide at its actual reported value or its MDC, then the analysis was repeated using only the actual reported values of the detected nuclides. Those nuclides included in the dose analysis at their MDC values were shown to contribute less than 10 percent of the annual dose from the pathway analyzed.

The soil profile given in Table 2-11 is used for both surface (within 15 cm of the surface) and deep (below 15 cm of the surface) soils. The soil fractions were decayed to 1/1/2004.

For additional discussion on soil samples and nuclide fraction see l Attachment 2I. l Table 2-11 Soil Nuclide Fractions Nuclide Fraction as of 1/2004 H-3 0.053 Ni-63 0.048 Co-60 0.009 Cs-137 0.890 6

Regarding buried and embedded piping and its impact on soil contamination, the most significant of l buried/embedded piping within the industrial area are the HPCI and LPCI lines. These contained the l same fluid as the RWST and would be well represented by the RWST and the subsequest RWST l related soil samples used in part of the soil nuclide fraction. l

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d. Groundwater and Surface Water Samples were taken of the groundwater (containment foundation sump) and the surface water sources (fire pond and reflecting pond). The samples were analyzed for gamma emitters and HTDs. Since the samples contained relatively low levels of residual activity, long count times were l used to achieve low MDAs. The only nuclide detected in either source of water was tritium. The surface water tritium is naturally occurring.

Additional information regarding background tritium in and around the l Maine Yankee site is provided in a comprehensive report on site l hydrogeology (Stratex, February 2002, Reference 2.7.19). l l

The February 2002 Stratex report (referenced above) summarized and l discussed radioactivity in site groundwater and its relationship to site l history regarding releases of contamination.7 In general, while relatively l low levels of Co-60 and Cs-137 have been sporadically detected in the l containment foundation sump and other site wells, the primary, l consistently detected nuclide is tritium. The nuclide fraction for l groundwater (used as an initial condition for the dose assessment) consists l of tritium only. See Section 6.6.6 for additional discussion, activity levels, l and the use of this nuclide fraction in the dose assessment. l l

An additional groundwater re-sampling program consisting of fifteen wells l was implemented in spring of 2002. The results of this effort, which l included the analysis of twelve of the fifteen well samples for hard to l detect nuclides, were reported in Maine Yankees letter to the NRC, l dated August 28, 2002 (Reference 2.7.20). This submittal included an l addendum to the February 2002 Stratex report (August 2002). This l sampling effort included not only the containment foundation sump but l also numerous wells in the industrial area as well as several new wells, as l recommended in the February 2002 Stratex report. (Additional l groundwater exploration of the Primary Auxillary Building PAB test pit l area, as recommended by Stratex in February 2002, was not pursued. See l discussion below.) l l

Consistent with prior well sampling in the industrial area, the results of l this site groundwater re-sampling effort showed relatively low levels of l groundwater contamination. Two wells reported relatively low levels of l either Co-60 and Cs-137. Tritium levels were above background in l several wells; however, they were consistent with previously detected l 7

Stratex, February 2002, Section 3.7 (LTP Reference 2.7.19). l

MYAPC License Termination Plan Page 2-61 Revision 3 October 15, 2002 concentrations and well within the conservative levels assumed for dose l modeling. Hard to detect analyses (including transuranics) detected no l other nuclides, also consistent with prior sampling. (See Reference l 2.7.20.) The nuclide fraction for both ground and surface water is given in l Table 2-12.

Special consideration and assessment was given to the isolated detection l (1999) of contamination in the PAB test pit, as discussed in the February l 2002 Stratex report. Additional study of the fate and transport of relevant l nuclides was performed by Stratex, supported by Brookhaven National l Laboratory (reported in the August 2002 Maine Yankee submittal to the l NRC). Based on the additional study, including consideration of recent l sampling of the test pit and the containment foundation sump and site l hydrogeology, Maine Yankee concluded that no additional field l investigations or groundwater exploration were necessary to further study l the fate and transport of the historical PAB test pit contamination. In that l the PAB test pit is a structure to remain post-decommissioning, it will l undergo any necessary remediation and final status surveys to demonstrate l compliance with surface contamination release criteria. (See Reference l 2.7.20.) l Samples from the containment foundation sump and the PAB test pit will l be routinely obtained and analyzed until the final status survey is l commenced for these two plant areas. See Section 6.6.6. Furthermore, as l noted in Section 6.6.6, future groundwater sampling data obtained prior to l unrestricted release will be considered for its impact on the dose l assessment. l Table 2-12 Ground and Surface Water Nuclide Fraction l Nuclide Fraction H-3 1.000

e. Forebay and Diffuser Contaminated Media l A detailed discussion of the characterization of the forebay and diffuser l system is provided in Attachment 2H. The characterization effort and l resulting nuclide fraction for forebay/diffuser media are summarized l below. l l

The forebay (and seal pit) characterization consisted of sampling efforts l that identified the following contaminated media: l

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1. Rock floors and walls of the forebay/seal pit, as well as a limited l amount of concrete surfaces at the northern and southern ends of l the forebay basin; l

2. Rip-rap, contaminated surfaces; l

3. Marine sediment deposited on the floors of the forebay/seal pit and l around the rip-rap; and l

4. Dike soil, i.e., that material beneath the rip-rap, interior to the l dike walls. l l

Sampling and assessment of the diffuser system identified two l contaminated media, namely, sediment entrained inside the diffuser l discharge piping and contaminated surface film deposited on the inside l surfaces of diffuser piping. This surface contamination was noted to be l very similar to that on the rip-rap covering the interior forebay dike walls. l l

As the results of several sampling campaigns (including diving l operations), each of the above media were sampled, analyzed, and l evaluated regarding nuclides present, activity levels, and relative fractions. l The evaluation included three sets of sediment samples analyzed for HTD l nuclides. The overall assessment concluded that a single nuclide fraction l was appropriate and conservative for application to these media. The l nuclide fraction for forebay and diffuser related media is presented in l Table 2-13. See Attachment 2H for additional discussion on the principal l construction features of the forebay and diffuser system, the sampling l campaigns, results, and conclusions. See also EC 041-01 for supporting l technical bases and analyses. l Table 2-13 Forebay/Diffuser Material Nuclide Fractions l Nuclide Fraction (as of 1/1/2004)

Fe-55 0.165 Ni-63 0.233 Co-60 0.567 Sb-125 0.005 Cs-137 0.030

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f. Future Sampling l l

The radionuclide profiles for contaminated concrete, activated concrete, soil, ground water, surface water, and sediment listed in Tables 2-7 and 2- l 8, 2-9, 2-11, 2-12, and 2-13 respectively, were determined using l representative data. These profile results do not rule out the possibility of taking additional samples of these media as decommissioning progresses and as conditions warrant.

Note: If radionuclide profiles are revised, the revised profiles will be provided to the NRC and the State of Maine at least 30 days prior to their use.

2.5.4 Background Determination The residual radioactivity of a survey unit may be compared directly to the DCGL; however, some survey units will contain one or more radionuclides which are also contained in background. In order to identify and evaluate those radionuclides, background areas have been established which contain only background levels of the radionuclides of interest. These background areas were chosen because they were similar in physical, chemical, geological and biological characteristics to the survey units.

a. Soils Soil samples were taken (ICS) from the non-impacted areas and analyzed l in order to establish general soil background levels. If background reference area measurements are required for the Final Survey Program, the reference area measurements will be collected in accordance with the methods described in Section 5 and the applicable approved procedures.

The samples showed mean Cs-137 levels of 0.2 to 0.5 pCi/g depending on whether the soil had been disturbed or not. The more undisturbed the soil is, the higher the background Cs-137 may be (e.g. Knight Cemetery, Eaton Farm, values reported in Attachments 2A & 2B). The naturally-occurring l uranium isotopes (U-234, U-235, and U-238) were present in expected amounts. Uranium is naturally occurring, not plant derived. These nuclides are not included in the Soil Mixture Nuclide Fraction listed in Table 2-11 above. Sr-90 was not detected at or above a MDC of 0.4 pCi/g.

MYAPC License Termination Plan Page 2-64 Revision 3 October 15, 2002

b. Structures Background measurements were taken on structural materials during initial characterization (ICS) in order to estimate the contribution of l background activity to the total measurement value. The same types of detectors will be used for FSS as were used during both ICS and CCS. l Background values for structural materials using these detectors are shown in Table 2-14.

Table 2-14 Structural Material Backgrounds Background Counts per Minute (reflects beta count rate) l Materials 43-68 Proportional SHP-360 G-M Pancake Detector - 126 cm2 Detector - 15.5 cm2 Painted Cinder Block 296** 70**

Wood 301** 57**

Ambient 319** 65**

Steel 277* 46*

Carpet 339** 68**

Floor Tile 359* 62*

Ceiling Tile 439* 73*

Bare Cinder Block 394** 79**

Painted Concrete 392* 74*

Bare Concrete 433* 76*

Asphalt 559* 99*

Granite 566** 128**

Porcelain 607** 116**

Brick 716* 118*

• Average of twenty-five one minute static counts taken in the scaler mode.

• • Average of ten one minute static counts taken in the scaler mode.

The 43-68 proportional detector will generally be used for surface contamination measurements because of its sensitivity, larger detection area and lower MDC. SHP-360 will only be used where a measurement can not be taken with a 43-68 detector.

MYAPC License Termination Plan Page 2-65 Revision 3 October 15, 2002 2.6 Summary 2.6.1 Impact Of Characterization Data On Decontamination And Decommissioning Characterization data (both ICS and CCS) confirmed what was known about the l MY site in terms of the level and extent of radioactive contamination. A major portion (700 acres) of the site met the classification of non-impacted. Primary systems and structures were found to be contaminated to expected levels. Non-RA systems and structures were found to be free of contamination except as previously stated.

There were minimal or no changes in either waste volumes or waste activity values following the performance of site characterization.

The data compiled are sufficient to project schedules and waste volumes, evaluate decontamination techniques, perform dose assessments and evaluate any safety or health issues affecting workers on site.

The HSA and characterization measurement results (ICS and CCS) are sufficient l to meet the objectives listed in Section 2.1 and demonstrate compliance with the guidance contained in Regulatory Guide 1.179 and NUREG-1700. The more than 19,000 measurements provide sufficient data to determine the radiological status of the site and facility as well as identify the location and extent of contamination outside the RA. The radionuclide analyses performed were sufficient to estimate the source term and isotopic mixture (based on the achieved standard deviation of the data). The analysis results also provide sufficient information to support dismantlement, radioactive waste disposal, decommissioning cost estimates and remediation decision making processes. The source term information was also suitable for instrument selection. The radiological data were acceptable to develop the necessary quality assurance methods for sample collection and analysis. The data obtained during characterization (ICS and CCS) support dose l assessment and FSS design.

MYAPC License Termination Plan Page 2-66 Revision 3 October 15, 2002 2.7 References 2.7.1 NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual, (MARSSIM), Revision 1 (June 2001) l 2.7.2 10 CFR.50.75, Reporting and Recordkeeping for Decommissioning Planning.

2.7.3 Continuing Characterization (CCS) Plan (PMP 6.8). l 2.7.4 CCS Quality Control (PMP 6.8.4). l 2.7.5 Corrective Action Program 2.7.6 Document Control Program (0-17-1). l 2.7.7 Radiation Protection Performance Assessment Program (PMP 6.0.8). l 2.7.8 Selection, Training and Qualification of Radiation Protection Personnel, (PMP 6.9). l 2.7.9 Maine Yankee Atomic Power Co. (MY), RCRA Quality Assurance Project l Plan for Maine Yankee Decommissioning Project, Revision 1. l (June 28, 2001) l 2.7.10 NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions.(June 1998) l 2.7.11 NUREG-1700, Standard Review Plan for Evaluating Nuclear Power Reactor License Termination Plans. (April 2000) l 2.7.12 Regulatory Guide 1.179, Standard Format and Content of License Termination Plans for Nuclear Reactors. (January 1999) l 2.7.13 NUREG/CR-3474, Long-Lived Activation Products in Reactor Materials.

2.7.14 GTS Duratek, Characterization Survey Report for the Maine Yankee Atomic Power Plant, Volumes 1-9, 1998 (ICS). l 2.7.15 Dr. Chabot letter to P. Dostie, dated November 12, 1998, discussing determination of MDC

MYAPC License Termination Plan Page 2-67 Revision 3 October 15, 2002 2.7.16 Maine Yankee letter to the NRC, MN-02-002, dated January 16, 2002, l transmitting special report from the Technical Issue Resolution Process, l entitled Transuranic and Other Hard to Detect Radionuclides in Maine l Yankee Sample Media. l l

2.7.17 NRC letter to Maine Yankee, dated July 30, 2002, Issuance of Amendment l No. 167, license amendment approving partial release of site lands. l l

2.7.18 Maine Yankee Engineering Calculation, EC-041-01 (MY), Revision 0 l l

2.7.19 Maine Yankee letter to the NRC, MN-02-010, dated February 20, 2002, l Maine Yankee Response to NRC RAI #16 (dated December 18, 2001) l Addressing Site Hydrogeology, (included submittal of Stratex, LLC, l report, Site Hydrogeology Description, Maine Yankee, Wiscasset, Maine, l February 2002). l l

2.7.20 Maine Yankee letter to the NRC, MN-02-037, dated August 28, 2002, l Maine Yankee Addendum Report Regarding Site Hydrogeology, l (including Stratex, LLC, report Site Hydrogeology Addendum, Maine l Yankee, Wiscasset, Maine, August 2002). l 2.7.21 Maine Yankee letter to the NRC, MN-02-011, dated March13, 2002, l Response to NRC Request(s) for Additional Information for the Maine l Yankee Atomic Power Station LTP l 2.7.22 MYAPC Historical Site Assessment (HSA), transmitted by MN-01-038 l dated October 1, 2001 l 2.7.23 Maine Yankee letter to the NRC, MN-02-015, dated April 11, 2002, l Revised Maine Yankee Response to NRC RAI #5 (dated December 18, l 2001) - Supplementary Historical Site Assessment (HSA) Data l l

2.7.24 Maine Yankee letter to the NRC, MN-02-045, dated October 8, 2002, l Minor Changes to Maine Yankee Responses to NRC Request for l Additional Information l

MYAPC License Termination Plan Attachment 2A Revision 3 Page 1 of 11 October 15, 2002 ATTACHMENT 2A Non-Impacted Area Assessment

MYAPC License Termination Plan Attachment 2A Revision 3 Page 2 of 11 October 15, 2002 ASSESSMENT OF THE MY SITE WEST AND NORTH OF BAILEY POINT FOR CLASSIFICATION AS NON-IMPACTED 2A. 1 Introduction One aspect of the FSS Plan is the proper classification of areas within the site. Areas must be classified as either: Impacted, Class1, Class 2, or Class 3; or Non-impacted. Non-impacted areas are defined in NUREG-1575 (MARSSIM) as areas that have no reasonable potential for residual contamination, no radiological impact from site operations and are typically identified during the Historical Site Assessment. The MY Historical Site Assessment (HSA) did not classify any areas within the site but it did provide data which could be used in conjunction with other information to classify areas. The HSA was not and will not be solely relied upon to make any classification, remediation or survey decision. The source term was well understood through previous Part 61 analysis. The potential pathways for this source term to potentially affect any offsite areas are well understood, described in the Off Site Dose Calculation Manual and monitored on a routine basis.

2A.2 Area Description Approximately 641 acres of the MY site are found to the West of Bailey Cove, North of the l access road (Ferry Road) and bounded by Back River to the east. The land is generally located beyond the 2000 foot exclusion zone established under the requirements of 10 CFR 100. As such, the area has been open and accessible to the general public and is bounded by residential land owners.

The referenced area consists of open fields, woodland and some shoreline property which has been uninhabited and unfarmed since plant construction started in 1968. The geology and hydrology of the area has been described in detail in the MY FSAR and is physically similar to the operating area of the site itself except for there being little or no surface soil disturbance (except for the ash pit and the ash pit access road). Structures in the area generally predate the construction of the plant.

The meteorology of the area has been characterized in detail in terms of annual precipitation, prevailing winds and stability class. Average annual precipitation exceeds the US average.

Prevailing winds are from the South but a sea breeze blows East to West.

2A.3 Historical Site Assessment The land areas under consideration are approximately 0.25 miles or more from the Reactor Building and process buildings. No radioactive material was used or stored beyond the peninsula of Bailey Point. License restrictions and administrative controls have been in place since power

MYAPC License Termination Plan Attachment 2A Revision 3 Page 3 of 11 October 15, 2002 operations began in 1972 to prevent unauthorized removal of radioactive material from the owner controlled area. Planned offsite releases of radioactive material were limited to the permitted effluent releases (which were kept ALARA by process controls) and radioactive solid waste which was shipped to licensed burial sites. The HSA documented approximately 120 actual or potential events involving unplanned releases of radioactive material or hazardous material during the 25 year operating history of the plant. Of these events, about two thirds involved or potentially involved radioactive material. Based on a review of the documentation assembled in the HSA, none of these events would have resulted in residual contamination of the area under consideration. Therefore, there is no reasonable potential for residual contamination in the area.

2A.4 Radiological Environmental Monitoring Program A Radiological Environmental Monitoring Program (REMP) was instituted prior to operation of the plant and continues to the present time. Environmental measurements taken have included thousands of gamma dose rates, hundreds of air and water samples, and hundreds of food stuff and surface vegetation samples. The key indicators of radiological impact in the area of concern are TLD measurements, air samples, water samples, vegetation samples, food crop samples and sediment samples.

TLD measurements have shown no difference in dose rates between the area under discussion and the control areas further from the site. Bailey Farm well water had slightly lower tritium levels on average than the water supplies in the Wiscasset area. Precipitation tritium levels at local sampling stations (Eaton and Bailey Farms) were similar to the control station levels.

Fruits and vegetables sampled at the Bailey Farm showed the presence of only K-40 and fallout-produced Cs-137. Grasses sampled at the Eaton and Bailey Farms showed only natural K-40 and fallout-produced nuclides during periods of atmospheric testing. Initial soil samples had Cs-137 at levels consistent with published values for fallout activity. Samples taken during the intervening period had Cs-137 levels consistent with that which should have resulted from the decay of the initial 1970 sample activity. No radionuclides of plant origin were detected in these areas.

2A.5 Special Surveys And Reports The HSA and other sources document samples (or measurements) of radiation and radioactive materials taken in the area in question. Pressurized ion chamber readings, TLD measurements, soil samples and even a fly over dose rate survey have documented radiation levels in the area similar to, or slightly less than, those measured in pre-operational surveys. The slight decline in levels is likely due to decreased levels of fallout-produced Cs-137 (Aerial Radiation Measurement Study, 1974 and University of Maine, 1974 and 1997). Some anomalous Cs data for Knight Cemetery, Eaton Farm and Foxbird Island can be understood in light of normal spacial variability in activity related to differences in sampling locations and the relatively undisturbed nature on some of these locations. Table 2A-7, Alternate Table of Cs-137

MYAPC License Termination Plan Attachment 2A Revision 3 Page 4 of 11 October 15, 2002 Activity, shows very consistent results and the impact of decay when 1970 and 1997 data are presented. It is not surprising that some of the Cs data increased with time up to 1974 since atomic weapon atmospheric testing was still being conducted up to 1974.

Based on NUREG-1575 guidance, classification of an area as not impacted can be made solely on the Historical Site Assessment. Rather than rely solely on the HSA, the area in question was subjected to site characterization surveys. During 1997 and 1998, GTS performed site characterization measurements in the area which included gamma dose rates determined by pressurized ion chamber and micro R meter, soil samples and drive around surveys using a vehicle-mounted 1.5"x 3"x 33" scintillation detector. The characterization surveys (PIC and drive around) in the area produced one area with an elevated radiation level. Upon investigation, the elevated reading was found to be due to local increase in naturally occurring radiation. Approximately 150 soil samples taken throughout the area showed only background levels of radioactive material in quantities slightly less than those reported in the 1972 pre-operational studies in this area which is consistent with the decay of the fallout-produced activity.

2A.6 Backlands Report l l

On August 16, 2002, Maine Yankee submitted an application1 for amendment to its license to l release these backlands from the jurisdiction of the license. This application was supplemented2 l on November 19 2001. In the supporting justification attached to the application, Maine Yankee l reviewed the soil sample Cs-137 results of the Initial Characterization Survey (ICS) to determine l if the residual radioactivity, if any, in the backlands is indistinguishable from background and l thereby support the classification of non-impacted. l l

Demonstrating indistinguishability from background employs MARSSIM Scenario B. In l Scenario B, the null hypothesis is that the survey unit meets the release criterion l (indistinguishable from background). Under Scenario B, the comparison of measurements in the l reference area and survey unit is made using two nonparametric statistical tests: the Wilcoxon l Rank Sum (WRS) test and the Quantile test. The WRS and Quantile tests are both used because l each test detects different residual contamination patterns in the survey units. Because two tests l are used, the Type I error rate, ", (normally set at 0.05) is halved, and set at 0.025, for the l individual tests. Using the NUREG-1505 recommended " of 0.025 allows for the use of the look- l up tables in NUREG-1505, for r and k values used in the Quantile test. l l

The WRS test is designed to determine whether or not a degree of residual radioactivity remains l 1

Maine Yankee Letter to USNRC dated August 16, 2001, Early Release of Backlands (Combined), l Proposed Change No. 211, Supplement No. 1, (MN-01-034) l 2

Maine Yankee Letter to USNRC dated November 19, 2001, Early Release of Backlands (Combined), l Proposed Change No. 211, Supplement No. 2, (MN-01-044) l

MYAPC License Termination Plan Attachment 2A Revision 3 Page 5 of 11 October 15, 2002 uniform throughout the survey unit. The Quantile test is designed to detect a patchy l contamination pattern. l Table 2A-8 contains the soil sample Cs-137 results for the background reference area. The l background reference area consisted of area surrounding the Marrymeeting Airfield located l approximately 10 miles from the site and was representative of site characteristics. The Kruskal- l Wallis test was used to confirm that there was no significant difference in the mean background l concentrations among potential reference areas. l l

Table 2A-9 summarizes the results of the soil sample Cs-137 results for the backlands areas and l compares them to the results for the background reference area. For each of the backlands areas, l the results of the WRS test, where applicable3, and the Quantile test successfully demonstrated l that the residual radioactivity, if any, in the areas was indistinguishable from background. l 2A.7 Conclusion l Based on the evaluation of the historical use of the area, the lack of use or storage of radioactive material in the area, the Historical Site Assessment findings, the REMP results, the results of the l site characterization surveys, and the demonstration of indistinguishability from background l described in the Backlands Report, the area to the West of Bailey Cove and North of Ferry Road l within the land owned by MY has been classified as non-impacted.

The area lends itself to use as a background reference area for soil samples and may be used as such during the FSS. Random sampling of soil in order to establish background activities may be l performed in this reference area, but no systematic sampling as required by MARSSIM for impacted areas will be performed. l 3

For area R-1500 Ash Rd. Rubble Piles, the maximum Cs-137 reading was less than the value known l as the Upper Boundary of the Grey Region; therefore, the application of the WRS test was not l necessary to demonstrate indistinguishability from background. l

MYAPC License Termination Plan Attachment 2A Revision 3 Page 6 of 11 October 15, 2002 Table 2A-1 RADIOLOGICAL ENVIRONMENTAL DATA TLD DATA (Mean Value in µR/hr)

Data Source Inner Ring Outer Ring Control Period; #

locations MY 11.8 12.0 11.9 1970-1972 n=9 MY 7.1 7.4 7.8 1990-1997 n=28 Univ. of Maine 8.2 8.6 9.3 1971-1996 n=87 Table 2A-2 Pressurized Ion Chamber Data (µR/hr)

Data Source Location 1971 1996 1998 Univ. of Maine Bailey House 9.5 8.8 Univ. of Maine Eaton Farm 9.5 9.3 Univ. of Maine Westport 11.4 9.1 Univ. of Maine Knight Cemetery 8.7 Univ. of Maine Long Ledge 9.0 GTS Merrymeeting Mean=8.2 Airfield Range: 7.2-9.8 n=300

MYAPC License Termination Plan Attachment 2A Revision 3 Page 7 of 11 October 15, 2002 Table 2A-3 Soil Cs-137 (pCi/g)

Sample 1970 1972 1974 1996 1997 GTS Location MY MY MY MY Characterizati on Bailey House 0.64 1.67 1.8 0.4 0.21; n=30 Bath 0.66 Dresden 0.58 Eaton Farm 0.53 0.87 2.5 0.09 0.45; n=60 l Edgecomb 0.48 Foxbird 0.35 0.48 Knight 4.96 2.42 Cemetery Long Ledge 0.80 0.38 Harrisons 0.52 Mason 0.68 Station Montsweag 0.42 Dam Westport 0.56 1.11 1.03 North of 0.39; n=60 Ferry Road Merrymeeting 0.42; n=60 l Airfield Shoreline 0.20; n=30 Mean Value 0.56 1.63 2.15 0.80 0.32

MYAPC License Termination Plan Attachment 2A Revision 3 Page 8 of 11 October 15, 2002 Table 2A-4 Surface & Well Water Data Sample Location (Mean H-3 pCi/L) 1977-1984 Bailey House 235 Montsweag Dam 276 Morse Well 187 Biscay Pond 297 Wiscasset Reservoir 278 Table 2A-5 Precipitation Data Sample Location (Mean H-3 pCi/L) 1977-1982 Bailey House 416 Eaton Farm 417 Westport 422 Dresden 397

MYAPC License Termination Plan Attachment 2A Revision 3 Page 9 of 11 October 15, 2002 Table 2A-6 Air Particulate Data (Mean Gross Beta Activity, pCi/m 3 )

MY Pre-Operational Data 1970 0.12 1971 0.12 1972 Jan-Jun Zone I=0.06, Zone II=0.07 Univ. of Maine 1981-1997 MY 1988-1998 Wiscasset 0.02* Montsweag 0.021 Augusta 0.02* Bailey House 0.020 Mason Station 0.020 Westport 0.021 Dresden 0.022

• Values estimated by graph. Individual data not available.

References:

MY data were taken from the REMP Reports for the time periods listed or the GTS Characterization Report.

University of Maine data were taken from A Radiological Survey of the Area Surrounding the MY Nuclear Plant, March 1997.

MYAPC License Termination Plan Attachment 2A Revision 3 Page 10 of 11 October 15, 2002 Table 2A-7 Alternate Table of Cs-137 Activity Soil Cs-137 (pCi/g)

Sample Location 1970 MY 1997 GTS Characterization Bailey House 0.64 0.21; n=30 Bath 0.66 Dresden 0.58 Eaton Farm 0.53 0.45; n=60 l Edgecomb 0.48 Harrisons 0.52 Mason Station 0.68 Montsweag Dam 0.42 Westport 0.56 North of Ferry Road 0.39; n=60 Shoreline 0.20; n=30 Mean Value 0.56 0.32 Table 2A-8 l Reference Area Soil Sample Cs-137 Results pCi/g l Reference Areas - Merrymeeting Airfield Mean Std. Dev. Number of ll (Ave.) (1 F) Samples l Combined (wood, open & scrub) 0.42 0.21 50 l Wood Land 0.47 0.24 10 l Open Land (Hay Field) 0.38 0.12 30 l Scrub Land 0.48 0.34 10 l

MYAPC License Termination Plan Attachment 2A Revision 3 Page 11 of 11 October 15, 2002 Table 2A-9 Soil Sample Cs-137 Results l Area Description Minimum Median Average Maximum Number l Cs-137 Cs-137 Cs-137 Cs-137 of l pCi/g pCi/g pCi/g pCi/g Measurements l Reference Area R2200 0.09 0.38 0.42 1.40 50 l Survey Unit R1500* Ash Rd. Rubble Piles 0.02 0.06 0.07 0.21 30 l Survey Unit R1600 Eaton Farm 0.05 0.39 0.45 1.43 60 l Survey Unit R1700 North of Old Ferry R. 0.04 0.30 0.39 1.55 60 l

• Disturbed open land area within R1700 North of Ferry Rd. l

MYAPC License Termination Plan Attachment 2B Revision 3 Page 1 of 18 October 15, 2002 ATTACHMENT 2B Characterization Data

MYAPC License Termination Plan Attachment 2B Revision 3 Page 2 of 18 October 15, 2002 Table 2B-1 Group A Radiological Characterization Results For Affected Structures and Surfaces Direct Beta DPM/100 cm2 Removable Beta DPM/100 cm2 Removable Alpha DPM/100 cm2 Exposure Rate MicroR/hr Package Mean Maximum Std. Dev. Mean Maximum Std. Dev. Mean Maximum Std. Dev. Mean Maximum Std. Dev.

(MDC) (MDC (MDC Minimum

) ) Detectable Exp Rate (MDER)*

A0100 81,976 1,970,974 259,134.5 296 4,282 598.7 0.0 2.4 0.5 2,375 4,065 816 Cont.El -2 Ft (30,453) (33) (8.4) (15)

A0200 62,970 2,238,614 247,399.2 2,388 128,734 13,577.2 0.7 7.3 1.6 887 1,961 463 Cont. El 20 Ft (16,277) (35) (9.7) (15)

A0300 38,444 345,960 55,889.2 1,469 31,054 3245.7 0.2 5.8 1.1 499.5 2,408 387.5 Cont. El 46 Ft (16,058) (33) (8.7) (15)

A0400 6,815 312,939 32,365.4 38.4 879 106.2 -0.1 1.8 0.6 706.6 2,901 649.7 Fuel Bldg El 21 Ft (12,436) (32) (8.5) (15)

A0500 438 2,659 792.6 4.9 20.3 7.0 0.1 3.9 1.0 14.0 14.6 0.9 DWST (2,322) (32) (8.4) (15)

A0600 1,106 32,328 7513.5 5.2 32.3 8.0 -0.1 3.9 0.7 1,100 3,477 827 PAB El 11 Ft (13,168) (32) (8.5) (15)

A0700 460 25,000 4655.1 5.9 51.5 9.7 -0.2 1.8 0.3 581 4,068 950 PAB El 21 Ft (15,837) (32) (7.7) (15)

A0800 508 14,073 2166.5 5.9 94.2 11.0 0.1 2.0 0.6 187 769 182 PAB El 36 Ft (18,042) (34) (7.0) (15)

A0900 699 18,955 2927.8 9.2 251 26.6 -0.6 3.9 0.6 42 501 78 RA Svc Bld (1,970) (34) (8.2) (15)

A1100 852 74,216 6023.3 0.3 35.8 7.0 0.1 4.1 0.8 334 3,563 752 LLWSB (17,886) (38) (8.1) (15)

A1200 73,939 2,233,580 379,578.7 128.7 2,073 323.1 -0.1 1.8 0.6 2,162 12,389 2,864 RCA Storage (26,286) (37) (8.6) (15)

MYAPC License Termination Plan Attachment 2B Revision 3 Page 3 of 18 October 15, 2002 Table 2B-1 Group A Radiological Characterization Results For Affected Structures and Surfaces Direct Beta DPM/100 cm2 Removable Beta DPM/100 cm2 Removable Alpha DPM/100 cm2 Exposure Rate MicroR/hr Package Mean Maximum Std. Dev. Mean Maximum Std. Dev. Mean Maximum Std. Dev. Mean Maximum Std. Dev.

(MDC) (MDC (MDC Minimum

) ) Detectable Exp Rate (MDER)*

A1300 27.5 720.5 255.1 4.9 19.8 7.6 -0.1 1.9 0.5 27.1 122.7 33.7 Equip Hatch (600) (35) (7.8) (15)

A1400 350.2 6,758 1379.9 47.1 657.5 126.8 -0.2 1.9 0.3 47.5 180.2 41.2 Pers Hatch (2198) (35) (7.8) (15)

A1500 214.9 3,678 734.3 4.4 23.5 7.7 -0.2 3.9 0.6 9.4 14.0 2.6 Mech Pen (661) (38) (8.4) (15)

A1600 -138.0 557.1 269.7 1.9 18.2 6.9 0.0 1.8 0.6 12.7 14.0 1.2 Elec Pen (654) (37) (7.7) (15)

A1700 83,249 4,968,088 431,253.4 177.5 19,727 1445.2 0.0 2.0 0.4 1,598 9,041 2,124 Spray Bld (24,797) (37) (7.2) (15)

A1800 147.5 1,278 422.4 2.3 36.6 11.3 -0.1 1.8 0.5 18.9 34.9 7.1 Aux Feed (2,019) (37) (7.7) (15)

Pump A1900 130.6 2,563 725.3 0.6 24.6 7.0 -0.1 1.8 0.6 90.6 182.9 45.9 HV-9 (6318) (36) (8.2) (15)

A2100 3,602 54,719 13,158.9 2.7 72.4 13.5 0.0 1.8 0.7 687.5 1,078.4 374.0 RWST (21,587) (38) (8.4) (15)

A2200 7,269 43,189 10,833.4 7.1 73.2 16.9 -0.1 1.8 0.6 667.6 1,197 246.6 BWST (21,255) (36) (8.2) (15)

A2300 668 3,258 942.1 5.8 27.4 7.1 0.1 1.8 0.8 N/A N/A N/A PWST (2,780) (32) (8.4)

A2400 955.5 4,300 1062.8 3.5 30.7 7.3 0.4 5.8 1.3 N/A N/A N/A Test Tks (1438) (36) (8.2)

• NOTE: MDER values are for the instrument in a low background area.

MYAPC License Termination Plan Attachment 2B Revision 3 Page 4 of 18 October 15, 2002 Table 2B-2 Group B Unaffected Structures and Surfaces, Including Structural Background Survey Direct Beta DPM/100 cm2 Removable Beta DPM/100 cm2 Removable Alpha DPM/100 cm2 Exposure Rate microR/hr Package Mean Max Std. Mean Max Std. Dev. Mean Max Std. Dev. Mean Max Std.

(MDC) Dev. (MDC) (MDC) Minimum Dev.

Detectable Exp Rate (MDER)*

B0100 26.7 653.7 246.9 3.5 19.1 4.8 -0.3 4.8 0.9 9.0 15.2 1.9 Turb El 61Ft (636) (17) (7.6) (15)

B0200 215.8 1054.2 384.1 4.1 25.8 5.4 -0.5 2.0 0.7 10.2 12.5 1.1 Control Rm (7.6)

(Old) (616) (16) (15)

B0300 -91.0 552.5 299.7 1.9 11.7 4.8 -0.2 2.1 0.9 12.2 14.9 2.0 MCC (701) (17) (7.3) (15)

B0400 10.1 840.1 351.2 2.6 18.4 5.3 -0.6 0.7 0.4 11.2 12.8 1.6 Fire Pmp (610) (32) (8.2) (15)

B0500 62.1 8613.8 752.2 2.8 203.4 15.8 -0.4 2.1 0.7 8.6 17.3 2.8 Turb El 21Ft (649) (17) (7.3) (15)

B0600 48.2 2031.4 332.9 2.9 30.0 6.1 -0.1 3.5 0.9 6.3 13.7 2.9 Turb El 39 Ft (603) (17) (7.3) (15)

B0700 80.0 1621.5 411.1 2.8 19.9 5.0 -0.1 2.4 0.7 12.5 26.0 3.5 Svc. Bld.

Non-RCA (821) (32) (8.4) (15)

B0800 -82.7 451.4 286.0 5.5 19.9 6.1 -0.2 0.9 0.5 8.4 9.9 0.8 FOSB (587) (16) (6.7) (15)

B0900 -176.9 411.9 209.8 4.3 19.9 5.6 -0.1 0.9 0.6 10.8 13.1 1.6 EDGs (683) (16) (6.7) (15)

B1000 183.4 1309.7 492.6 3.4 16.5 5.9 -0.2 2.4 0.7 9.2 10.5 0.9 Aux Boiler (679) (16) (6.7) (15)

MYAPC License Termination Plan Attachment 2B Revision 3 Page 5 of 18 October 15, 2002 Table 2B-2 Group B Unaffected Structures and Surfaces, Including Structural Background Survey Direct Beta DPM/100 cm2 Removable Beta DPM/100 cm2 Removable Alpha DPM/100 cm2 Exposure Rate microR/hr Package Mean Max Std. Mean Max Std. Dev. Mean Max Std. Dev. Mean Max Std.

(MDC) Dev. (MDC) (MDC) Minimum Dev.

Detectable Exp Rate (MDER)*

B1100 -333.9 672.7 300.5 1.8 11.4 4.1 0.0 2.4 0.9 8.5 10.8 1.3 Circ Water (699) (16) (6.7) (15)

B1200 293.1 1628.2 431.9 4.3 14.8 5.1 0.0 2.4 0.9 13.3 15.2 1.5 Admin Bld (686) (16) (6.7) (15)

B1300 -146.3 1163.8 542.5 2.6 13.1 4.5 0.1 2.4 0.9 11.1 12.9 1.2 WART (666) (16) (6.7) (15)

B1400 295.3 1928.8 325.6 2.1 21.5 5.0 0.1 3.8 1.0 13.4 16.8 1.3 Info Ctr (678) (16) (6.7) (15)

B1500 96.1 539.0 212.4 0.6 19.4 5.2 -0.3 2.1 0.8 10.3 15.1 1.4 Warehse 2 (566) (18) (7.3) (15)

B1600 -13.5 708.2 256.1 1.6 17.7 4.8 -0.2 2.1 0.8 17.8 23.8 3.5 Trng Annex (657) (18) (7.3) (15)

B1700 129.4 952.9 279.5 -1.0 14.4 4.5 -0.4 3.5 0.7 14.2 23.2 3.3 Staff Bld (727) (18) (7.3) (15)

B1800 -39.8 341.9 176.6 0.1 9.3 4.6 -0.5 0.7 0.5 N/A N/A N/A Spare Gen Bld (548) (18) (7.3)

B1900 612.3 6523.7 1595.1 0.3 11.0 6.1 -0.4 0.7 0.6 9.4 16.1 3.6 Bailey House (682) (18) (7.3) (15 )

B2000 -96.6 306.5 187.3 1.1 9.3 4.6 -0.4 0.7 0.6 9.2 10.6 0.8 Bailey Barn (592) (18) (7.3) (15)

MYAPC License Termination Plan Attachment 2B Revision 3 Page 6 of 18 October 15, 2002 Table 2B-2 Group B Unaffected Structures and Surfaces, Including Structural Background Survey Direct Beta DPM/100 cm2 Removable Beta DPM/100 cm2 Removable Alpha DPM/100 cm2 Exposure Rate microR/hr Package Mean Max Std. Mean Max Std. Dev. Mean Max Std. Dev. Mean Max Std.

(MDC) Dev. (MDC) (MDC) Minimum Dev.

Detectable Exp Rate (MDER)*

B2100 8.7 610.4 240.7 0.2 7.6 4.3 -0.5 0.7 0.6 8.8 10.9 1.8 Lube Oil Storage (630) (18) (7.3) (15)

B2200 139.4 762.3 317.9 0.6 7.6 4.0 -0.5 0.7 0.5 8.0 9.0 0.9 Cold Shop (604) (18) (7.3) (15)

B2300 -23.4 275.3 195.1 0.5 21.3 5.0 -0.3 2.3 0.6 13.8 17.1 1.9 Cable Vault (632) (18) (6.9) (15)

B2400 19.2 575.6 359.6 3.8 18.0 6.7 -0.1 3.7 0.9 20.3 24.2 2.3 Staff Tunnel (779) (18) (6.9) (15)

• NOTE: MDER values are for the instrument in a low background area.

MYAPC License Termination Plan Attachment 2B Revision 3 Page 7 of 18 October 15, 2002 Table 2B-3 Group C Radiological Characterization Results For Affected Systems Direct Beta DPM/100 cm2 Removable Beta DPM/100 cm2 Removable Alpha Exposure Rate microR/hr Tritium DPM/100 cm2 DPM/ 100 cm2 Package Mean Max Std. Dev Mean Max Std. Dev. Mean Max Std. Mean Max Std. Dev. Mean (MDC) (MDC) (MDC) Dev. Minimum Detectable Exp Rate (MDER)*

C0100 N/A N/A N/A 77,858 300,000 126,236 1.5 8.0 3.7 1386 4161 1422.9 61.1 PASS (5000) (8.4) (15) (39)

C0200 N/A N/A N/A 2344 4073 2069.9 -0.3 -0.3 0.0 23,333 219,340 53,199 399.9 Waste Solid. (34) (8.4) ( 15) (39)

C0300 N/A N/A N/A 25,185 39,530 14,366.8 11.5 24.7 11.5 2593 22,862 4192 18.4 Contain. Spray (34) (8.4) (15) (39)

C0400 N/A N/A N/A 70,933 200,000 111,776 3.3 5.9 3.0 4416 34,960 6025 1377.8 ECCS (5000) (8.4) (15) (139)

C0500 N/A N/A N/A 76,000 180,000 91,476.8 N/A N/A N/A 4882 15,772 4112 23,617 RHR (5000) (15) (139)

C0600 N/A N/A N/A 50,585 140,000 77,438 -0.2 0.0 0.2 165,583 1,326,311 325,892 548 Pri. Vent & (5000) (8.4) (15) (39)

Drains C0700 N/A N/A N/A 13,693 20,000 6466.2 3.4 10.1 5.8 829,672 16,945,540 2,924,669 31.0 SFP Cooling (5000) (8.4) (15 ) (39)

C0800 N/A N/A N/A 3251 6470 2854.0 -0.3 -0.3 0.0 3295 23,554 4,999.5 5825 Waste Gas (34) (8.4) (15) (39)

C0900 N/A N/A N/A 213,333 360,000 128,582 N/A N/A N/A 41,636 376,269 59,187 82,468 Pzr. (5000) (15) (139)

C1100 N/A N/A N/A N/A N/A N/A N/A N/A N/A 53,580 181,323 34,275 N/A RCS (15)

MYAPC License Termination Plan Attachment 2B Revision 3 Page 8 of 18 October 15, 2002 Table 2B-3 Group C Radiological Characterization Results For Affected Systems Direct Beta DPM/100 cm2 Removable Beta DPM/100 cm2 Removable Alpha Exposure Rate microR/hr Tritium DPM/100 cm2 DPM/ 100 cm2 Package Mean Max Std. Dev Mean Max Std. Dev. Mean Max Std. Mean Max Std. Dev. Mean (MDC) (MDC) (MDC) Dev. Minimum Detectable Exp Rate (MDER)*

C1200 N/A N/A N/A 53,766 160,000 92,001.4 -0.2 0.0 0.2 1283 13,023 2078 19,515 Boron Recovery (5000) (8.4) (15) (39)

C1300 1907 3924.8 2074.1 29,197 112,370 47,511.3 8.8 34.9 14.8 41,446 884,946 127,708 1057 CVCS (1316) (7.8) (15) (139)

C1400 N/A N/A N/A 1078 1403 289.4 1.2 3.9 2.4 91,689 935,068 166,593 1187 Liq. Waste (35) (7.8) (15) (39)

C1500 N/A N/A N/A 1895 6002 2409.7 0.5 1.9 1.1 2059 10,306 2309 128.4 PAB Drains (35) (7.8) (15) (38)

C1600 5275 16,837 6185.7 52.8 194 72.0 -0.1 1.9 0.6 492.4 3546 1007 -17.6 PAB Vent (1144) (35) (7.8) (15) (38)

C1800 448,954 540,758 77,163.2 16,768 80,000 35,348.1 1.1 3.9 1.8 802.4 2275 653 -3.4 Contain. Vent (15,606) (5000) (7.8) (15) (38)

C1900 N/A N/A N/A 266,667 500,000 202,320 N/A N/A N/A 17,071 82,025 21,980 398.0 S/Gs (5000) (15) (139)

• NOTE: MDER values are for the instrument in a low background area.

MYAPC License Termination Plan Attachment 2B Revision 3 Page 9 of 18 October 15, 2002 Table 2B-4 -Group D Unaffected Systems Direct Beta DPM/100 cm2 Removable Beta DPM/100 cm2 Removable Alpha DPM/100 cm2 Exposure Rate microR/hr Package Mean Max Std. Dev. Mean Max Std. Mean Max Std. Dev. Mean Max Std. Dev.

(MDC) (MDC) Dev. (MDC) Minimum Detectable Exp Rate (MDER)

• D0100 66.7 2184.5 425.2 -0.5 14.6 5.1 -0.3 2.3 0.7 1.9 2.1 0.1 Condens. (15)

D0200 1250.8 26,046.3 4898.1 38.1 945.1 162.8 13.6 362.2 61.9 12.6 44.2 17.7 Water Treat. (1937) (16) (7.6) (15)

D0300 526.2 2638.6 767.7 6.7 29.2 6.9 0.4 9.1 2.3 4.5 7.1 1.6 Potable Water (1089) (16) (7.6) (15)

D0400 384.8 5657.1 1051.5 3.2 32.2 8.9 0.0 1.9 0.6 11.3 16.2 4.3 Sewer (1088) (36) (8.2) (15)

D0500 162.0 811.8 295.1 3.1 14.7 4.2 -0.1 5.1 0.9 3.7 17.2 5.1 Circ Water (587) (15) (6.9) (15)

D0600 38.0 1013.9 347.9 197.5 3133.7 658.5 -0.2 1.8 0.5 N/A N/A N/A Svc Water (1687) (37) (8.6)

D0700 -35.6 1114.7 240.2 2.4 20.6 5.2 0.2 2.5 0.9 N/A N/A N/A Fire Prot. (1257) (17) (6)

D0800 66.0 723.4 253.6 2.5 22.3 6.1 0.1 2.5 0.7 6.0 12.3 5.5 Lube Oil (1681) (17) (6) (15)

D0900 3677.5 104,589 14,456.3 27.0 685.2 95.1 0.4 6.8 1.4 N/A N/A N/A Comp. Air (6324) (17) (6)

D1000 446.0 2723.9 730.5 12.3 114.8 21.8 0.0 2.5 0.8 7.1 20.1 5.3 Aux Boiler (2606) (17) (6) (15)

D1100 270.8 2664.1 1067.4 9.2 47.5 11.1 0.3 2.5 1.0 35.0 66.8 44.9 S/G (1347) (17) (6) (15)

MYAPC License Termination Plan Attachment 2B Revision 3 Page 10 of 18 October 15, 2002 Table 2B-4 -Group D Unaffected Systems Direct Beta DPM/100 cm2 Removable Beta DPM/100 cm2 Removable Alpha DPM/100 cm2 Exposure Rate microR/hr Package Mean Max Std. Dev. Mean Max Std. Mean Max Std. Dev. Mean Max Std. Dev.

(MDC) (MDC) Dev. (MDC) Minimum Detectable Exp Rate (MDER)

• D1200 -9.2 4598.7 649.0 0.8 59.6 9.3 -0.3 2.2 0.7 N/A N/A N/A Main Steam (1002) (36) (8.2)

D1300 667.3 11,786.6 1963.4 1.9 19.4 6.5 0.0 2.0 0.5 162.8 435.1 218.9 Aux Steam (2382) (36) (8.2) (15)

D1400 -38.3 416.5 189.7 -0.9 20.9 6.3 -0.4 0.8 0.5 0.8 1.6 0.4 Turb Control (839) (19) (7.1) (15)

D1500 -216.5 64.1 139.9 -0.8 10.8 4.1 -0.5 0.8 0.5 N/A N/A N/A Steam Dump (677) (19) (7.1)

D1600 -0.3 453.9 160.8 -1.2 24.2 6.3 -0.4 2.2 0.6 2.0 5.4 2.2 Main Feed (640) (19) (7.1) (15)

D1700 -136.5 851.3 347.6 0.9 21.0 5.3 -0.3 3.6 0.8 N/A N/A N/A EFW (2414) (18) (7.1)

D1800 42.4 1864.3 323.3 -2.7 9.1 3.8 -0.4 2.2 0.6 0.9 1.3 0.4 Htr. Drain, (1182) (19) (7.1) (15)

Extract D1900 1168.0 21,644.3 6616.3 5.2 38.0 10.7 -0.1 2.0 0.3 10.1 12.8 2.0 Comp Cooling (4385) (36) (7.2) (15)

D2000 24.8 672.1 257.8 1.6 14.2 4.8 -0.3 2.2 0.8 N/A N/A N/A Vac Prim (1256) (18) (7.1)

D2100 107.5 1880.2 507.5 2.2 15.9 5.4 0.1 3.6 1.1 N/A N/A N/A Amertap (1200) (18) (7.1)

D2200 23.3 582.0 237.8 0.2 10.9 4.2 -0.5 0.8 0.5 N/A N/A N/A Sealing Steam (1067) (18) (7.1)

MYAPC License Termination Plan Attachment 2B Revision 3 Page 11 of 18 October 15, 2002 Table 2B-4 -Group D Unaffected Systems Direct Beta DPM/100 cm2 Removable Beta DPM/100 cm2 Removable Alpha DPM/100 cm2 Exposure Rate microR/hr Package Mean Max Std. Dev. Mean Max Std. Mean Max Std. Dev. Mean Max Std. Dev.

(MDC) (MDC) Dev. (MDC) Minimum Detectable Exp Rate (MDER)

• D2300 31.7 535.3 210.8 3.1 31.8 9.3 0.1 2.0 0.7 N/A N/A N/A Aux DG (645) (36) (7.2)

D2400 35.2 645.5 251.2 307.2 4861.3 995.8 0.3 6.0 1.4 N/A N/A N/A Chem Sample (1617) (35) (7.8)

D2500 132.2 594.8 260.3 -0.1 7.5 4.7 -0.4 0.8 0.6 N/A N/A N/A HP Drain (1048) (18) (7.1)

D2600 336.6 1257.1 400.1 3.7 12.9 3.9 0.6 3.9 1.3 N/A N/A N/A Envir (535) (14) (6.9)

D2700 74.3 643.3 276.3 5.2 32.8 8.5 0.3 2.2 1.1 8.0 8.0 0.0 Admin HVAC (789) (18) (7.1) (15)

D2800 156.2 627.8 256.9 0.6 10.9 4.7 -0.5 0.8 0.5 N/A N/A N/A Info Ctr Hvac (702) (18) (7.1)

D2900 142.4 445.4 161.5 4.6 33.1 5.9 0.3 3.9 0.9 N/A N/A N/A Turb HVAC (577) (14) (6.9)

D3000 262.9 1286.3 366.0 2.2 15.9 6.0 -0.1 2.2 0.9 N/A N/A N/A Staff HVAC (779) (18) (7.1)

D3100 5346.8 87,565.8 19,067.0 80.0 1445.0 247.1 0.6 5.9 1.3 22.4 51.4 17.4 Svc HVAC (1082) (14) (8.5) (15)

D3200 12,037.3 125,317 36,307.5 104.5 828.9 245.4 0.6 9.9 2.3 N/A N/A N/A H2/N2 (3059) (14) (8.5)

D3300 433.1 5800.9 1166.9 8.1 33.6 9.0 0.0 1.8 0.8 10.8 15.9 4.9 Turb Sumps (1091) (32) (8.4) (15)

MYAPC License Termination Plan Attachment 2B Revision 3 Page 12 of 18 October 15, 2002 Table 2B-4 -Group D Unaffected Systems Direct Beta DPM/100 cm2 Removable Beta DPM/100 cm2 Removable Alpha DPM/100 cm2 Exposure Rate microR/hr Package Mean Max Std. Dev. Mean Max Std. Mean Max Std. Dev. Mean Max Std. Dev.

(MDC) (MDC) Dev. (MDC) Minimum Detectable Exp Rate (MDER)

• D3400 457.0 3099.3 1300.0 7.1 27.4 8.7 0.1 6.0 1.3 N/A N/A N/A LLWSB (992) (32) (8.4)

• NOTE: MDER values are for the instrument in a low background area.

Table 2B-5 Group R Radiological Characterization Results For Affected and Unaffected Environs Exposure Rate microR/hr Package # # Mean Max # Mean Max Mean Maximum Std. Dev.

Sample Positive Co-60 Co-60 Positive Cs-137 Cs-137 s Co-60 pCi/g pCi/g Cs-137 pCi/g pCi/g R0100 58 23 0.62 3.29 55 10.99 156.0 N/A N/A N/A N/A N/A RA Yard West R0200 35 12 0.28 1.94 33 4.88 133.0 N/A N/A N/A N/A N/A Yard East R0300 7 4 4.09 11.2 6 0.33 0.53 N/A N/A N/A N/A N/A Roof Drains R0400 27 1 0.08 0.08 27 0.34 0.98 N/A N/A N/A N/A N/A Shoreline R0500 45 0 0 0 44 0.38 1.09 N/A N/A 13.27 19.83 1.49 Bailey Pt.

R0600 32 0 0 0 3 0.04 0.06 N/A N/A 11.92 13.68 0.63 Ball Field R0700 31 0 0 0 2 0.05 0.06 N/A N/A 11.99 14.52 1.05 Constr. Debris

MYAPC License Termination Plan Attachment 2B Revision 3 Page 13 of 18 October 15, 2002 Table 2B-5 Group R Radiological Characterization Results For Affected and Unaffected Environs Exposure Rate microR/hr Package # # Mean Max # Mean Max Mean Maximum Std. Dev.

Sample Positive Co-60 Co-60 Positive Cs-137 Cs-137 s Co-60 pCi/g pCi/g Cs-137 pCi/g pCi/g R0800 30 0 0 0 26 0.26 0.83 N/A N/A 17.9 33.87 4.2 Admin. Parking R0900 36 6 1.22 5.11 24 11.06 85.6 N/A N/A 25.85 77.71 16.8 BOP R1000 73 3 0.22 0.38 43 0.43 1.63 N/A N/A 11.48 42.76 4.97 Foxbird Is R1100 15 0 0 0 3 0.07 0.09 N/A N/A N/A N/A N/A Roof Drains R1200 30 0 0 0 5 0.10 0.13 N/A N/A N/A N/A N/A LLWSB Yard R1300 30 0 0 0 5 0.12 0.28 N/A N/A 12.92 31.2 3.68 ISFSI R1400 30 0 0 0 30 0.20 0.35 N/A N/A N/A N/A N/A Shorelines R1500 30 0 0 0 9 0.07 0.21 N/A N/A 11.34 12.63 0.63 l Ash Pit Rubble R1600 60 0 0 0 59 0.045 1.43 N/A N/A 12.07 17.8 2.06 l Eaton Farm Land R1700 60 0 0 0 50 0.39 1.55 N/A N/A 9.65 13.74 1.56 l Land North of Ferry Rd R1800 31 0 0 0 22 0.27 0.76 N/A N/A 10.63 14.57 1.31 Bailey Farm Land R1900 14 0 0 0 14 0.27 0.37 N/A N/A N/A N/A N/A Bailey Cove

MYAPC License Termination Plan Attachment 2B Revision 3 Page 14 of 18 October 15, 2002 Table 2B-5 Group R Radiological Characterization Results For Affected and Unaffected Environs Exposure Rate microR/hr Package # # Mean Max # Mean Max Mean Maximum Std. Dev.

Sample Positive Co-60 Co-60 Positive Cs-137 Cs-137 s Co-60 pCi/g pCi/g Cs-137 pCi/g pCi/g R2000 5 2 0.1 0.12 4 0.10 0.13 N/A N/A N/A N/A N/A Diffuser R2100 30 0 0 0 4 0.13 0.33 N/A N/A 8.41 10.62 1.33 Warehse Yard R2200 62 0 0 0 62 0.35 1.4 N/A N/A 11.37 13.59 1.26 Backgrnd* l R2300 16 1 0.14 0.14 15 0.35 0.81 N/A N/A 26.14 29.4 1.46 SFP Substation R2400 44 0 0 0 9 0.48 1.62 N/A N/A N/A N/A N/A IT Duplicates

• Includes twelve marine sediment samples taken the New Meadows River and the Damariscotta River. l

MYAPC License Termination Plan Attachment 2B Revision 3 Page 15 of 18 October 15, 2002 Table 2B-6 R2500 Investigation Package Package # Samples # Positive Co-60 Mean Co-60 pCi/g Max Co-60 pCi/g # Positive Cs-137 Mean Cs-137 pCi/g Max Cs-137 pCi/g R0500 8 3 11,218.5 33,600.0 7 0.13 0.21 Bailey Pt R0600 15 0 0 0 5 0.16 0.29 Ball Field R0700 40 0 0 0 3 0.04 0.06 Construction Debris R0800 15 0 0 0 14 0.17 0.33 Admin Parking Lot R1000 10 0 0 0 7 0.13 0.21 Foxbird Is R1300 10 2 0.43* 0.45* 4 0.07 0.12 l ISFSI R1600 5 0 0 0 2 0.27 0.29 Eaton Farm Land R1800 20 0 0 0 13 0.10 0.15 Bailey Farm Land

• Activity consisted, in part, of discrete particles l

MYAPC License Termination Plan Attachment 2B Revision 3 Page 16 of 18 October 15, 2002 Table 2B-7 R2501 Investigation Package Package # Samples # Positive Mean Co-60 pCi/g Max Co-60 pCi/g # Positive Cs-137 Mean Cs-137 pCi/g Max Cs-137 pCi/g R0900 41 16 0.12 0.49 41 17.1 145 BOP R1000 26 2 0.08 0.11 24 2.53 10.0 Foxbird Is.

R2500 27 0* 0* 0* 4 0.20 0.31 Contractors Parking 0 indicates less than MDC where MDC is #0.1 pCi/g for soil

MYAPC License Termination Plan Attachment 2B Revision 3 Page 17 of 18 October 15, 2002 Table 2B-8 Radiological Characterization Water Sample Results For Affected and Unaffected Environs, Including Environs Background Study Package Well/Catch Basin Tritium Activity Plant Derived Identification pCi/L Gamma Activity ?

R0100 203 1198 No 205 928 No 206 541 No BK-1 4023 No Chromate Well 914 No CTMT Foundation Sump 6812 No Average 2403 Package Well/Catch Basin Tritium Activity pCi/L Plant Derived Identification Gamma Activity ?

R0200 202 622 No 204 441 No MW100 788 No Average 617 Package Well/Catch Basin Tritium Activity Plant Derived Identification pCi/L Gamma Activity ?

R0300 6A 2005 No 7A 3266 No 7B 978 No 7E 2712 No Outfall #6 716 No Average 1935

MYAPC License Termination Plan Attachment 2B Revision 3 Page 18 of 18 October 15, 2002 Package Well/Catch Basin Tritium Activity Plant Derived Identification pCi/L Gamma Activity ?

R1100 9A 833 No 10A 815 No 11A 581 No Average 743 Package Well/Catch Basin Tritium MDA Plant Derived l Identification Activity pCi/L pCi/L Gamma Activity l

?

R2200 Eaton Farm Well 685 743 No l Bailey Farm Well -1689 3126 No l Days Ferry (private 1220 2255 No l well)

Average 635 2042 l Package Well/Catch Basin Tritium Activity Plant Derived Identification pCi/L Gamma Activity ?

R2400 North Transformer Sump 599 No Main Transformer Sump 842 No Groundwater Sump 756 No Edgecomb Average 73

MYAPC License Termination Plan Attachment 2C Revision 3 Page 1 of 5 October 15, 2002 ATTACHMENT 2C Summary of Continued Characterization Data

MYAPC License Termination Plan Attachment 2C Revision 3 Page 2 of 5 October 15, 2002 Table 2C-1 Group C Continued Characterization Results For Systems and Soils Package Direct Beta DPM/100 Isotopic Analysis Of Internals Co-60 Isotopic Analysis Of System Internals, cm2 (pCi/g) Cs-137 (pCi/g)

Mean Max Std. # Positives/ Mean Max Std. # Positives/ Mean Max Std.

(MDC) Dev. #Measurements Dev. #Measurements Dev.

CD0100 764 4923 1403 2/4 358 715 506 0/4 <MDC <MDC N/A Condensate (2351)

CD0200 499 1923 728 0/4 <MDC <MDC N/A 0/4 <MDC <MDC N/A Water Treatment (2351)

CD0600 -6819 - 872 3/3 2.92 5.44 2.31 0/3 <MDC <MDC N/A Svc. Water (5329) 3161 CD1900 106 1303 53 N/A N/A N/A N/A N/A N/A N/A N/A SCC (2086)

CD1900 3780 1331 3676 N/A N/A N/A N/A N/A N/A N/A N/A PCC (2351) 0 Package Soil Isotopic Analysis, Co-60 (pCi/g) Soil Isotopic Analysis, Cs-137 (pCi/g)

1. Positives/ Mean Max Std. #Positives/ Mean Max Std.

1. Samples Dev. # Samples Dev.

CR0200 N/A N/A N/A 0/25 <MDC <MDC N/A 12/25 0.19 0.32 0.09 Fuel Is.

Pagoda CR0500 N/A N/A N/A 0/11 <MDC <MDC N/A 4/11 0.14 0.21 0.06 Bailey Point CR1000 N/A N/A N/A 1/36 0.05 0.05 N/A 23/36 1.03 4.37 1.23 Foxbird Is.

CR1300 N/A N/A N/A 0/16 <MDC <MDC N/A 0/16 <MDC <MDC N/A Contr. Prk.

Lot MDCs ranged from: 0.1 - 0.4 pCi/g for soil samples 30 - 80 pCi/g for valve disks 30 pCi/smear for smear samples 0.02 - 0.2 pCi/g for pipe debris

MYAPC License Termination Plan Attachment 2C Revision 3 Page 3 of 5 October 15, 2002 Table 2C-2 Continued Characterization Results for Concrete Core Activity Concrete Core Samples (geometry corrected except as noted (1)) l Sample Net CPM Co-60 Cs-134 Cs-137 Eu-152 Eu-154 Area

1. 43-68 pCi/g pCi/g pCi/g pCi/g pCi/g (2) l 1-1A 49900 114 11 2038 Ctmt-2' 1-2A 132000 2545 125 5566 Ctmt-2' 1-3A 29800 354 9 307 Ctmt-2' 1-4A 82400 50 27 5616 Ctmt-2' 2-1A 1460 6 0.4 11 Ctmt 20' 2-2A 1230 3 1 16 Ctmt 20' 3-1A (1) 2920 190 39 172 285 Ctmt-32' 3-2A (1) 13300 307 37 359 290 35 Ctmt-32' 3-3A (1) 2460 157 28 36 280 33 Ctmt-32' l 4-1A 1270 1 0.4 14 Ctmt 46' 4-2A 18700 8 6 388 Ctmt 46' 4-3A 1960 3 1 35 Ctmt 46' 4-4A 2190 8 18 Ctmt 46' 4-5A 2920 6 0.6 29 Ctmt 46' 5-1A 2940 6 0.2 59 RCA 21' 5-2A 720 1 106 RCA 21' 5-3A 240 1 11 RCA 21' 5-4A 130 1.7 18 RCA 21' 5-5A 70 1 22 RCA 21' 5-6A 0 0 0 RCA 21' 5-7A 1090 37 63 RCA 21' 6-1A 18900 208 8 1030 PAB 11' 6-2A 130 0 4 PAB 11' 6-3A 1620 0 23 PAB 11'

MYAPC License Termination Plan Attachment 2C Revision 3 Page 4 of 5 October 15, 2002 Table 2C-2 Continued Characterization Results for Concrete Core Activity Concrete Core Samples (geometry corrected except as noted (1)) l Sample Net CPM Co-60 Cs-134 Cs-137 Eu-152 Eu-154 Area

1. 43-68 pCi/g pCi/g pCi/g pCi/g pCi/g (2) l 6-4A 0 0.4 2 PAB 11' 6-5A 0 0 0 PAB 11' 6-6A 0 0 0 PAB 11' 7-1A 630 1 7 PAB 21' 7-2A 0 0 0 PAB 21' 8-1A 410 0.3 13 Spray21' 8-2A 29610 35 809 Spray12' 8-3A 4380 4 62 Spray12' 8-4A 144000 152 3 4508 Spray12' 9-1A 190 2 38 Spray 4' 9-2A 340 2 3 Spray 4' 9-3A 110 0 2 Spray 4' 9-4A 140 6 6 Spray-6' 10-1A 40 0 4 Fuel 21' 10-2A 530 1 575 Fuel 21' 10-3A 550 2 14.7 Fuel 21' l 10-4A 8690 156 1186 Fuel 21' 11-1A 2200 0 64 Fuel 31' 11-2A 1380 0 20 Fuel 31' 12-1A 54426 935 9 636 Cntmt O/A Trench 12-2A 72326 931 9 535 Cntmt O/A Trench 12-3A 53151 374 22 3280 Cntmt El-2' 12-4A 12651 66 10 1179 Cntmt El-2' l

MYAPC License Termination Plan Attachment 2C Revision 3 Page 5 of 5 October 15, 2002 Table 2C-2 Continued Characterization Results for Concrete Core Activity Concrete Core Samples (geometry corrected except as noted (1)) l Sample Net CPM Co-60 Cs-134 Cs-137 Eu-152 Eu-154 Area

1. 43-68 pCi/g pCi/g pCi/g pCi/g pCi/g (2) l 12-5A 143651 664 56 11914 Cntmt El-2' 13-1A 1193 7 61 PAB El-11' 13-2A 14383 86 10 192 PAB El-11' l 13-3A 5273 52 2 47 PAB El-11' (1) Activation Samples (not geometry corrected) l (2) Net Count Rate. For additional discussion, see Attachment 2G. l Table 2C-3 Continued Characterization Results for Water and Sediment Samples CTMT Foundation H-3: 900 pCi/L Sump Gamma Spec and HTDs: No detectable Activity Reflecting Pond H-3: 600 to 960 pCi/L Gamma Spec: No Detectable Activity with 2E-9 µCi/ml MDA Forebay Sediment Fe-55: 13.6 pCi/g Composite (1) Ni-63: 8.9 pCi/g l Co-60: 31.7 pCi/g Sb-125: 0.4 pCi/g l Cs-137: 1.2 pCi/g (1) Results are from the 2000 composite forebay sediment sample. For l additional information, see Attachment 2H regarding forebay and diffuser l characterization. l

MYAPC License Termination Plan Revision 3 October 15, 2001 ATTACHMENT 2D Maine Yankee Site Characterization Locations of Radiological Survey Packages

MYAPC License Termination Plan Revision 3 October 15, 2002 N

R17 R15 R6 R7 R18 R9 R13 R16 R8 R21 R3,R1 R14 R12 1

R19 R4 R14 R4 14 R R14 R1 R10 4

R14 R14 ne ho re R20 li se tS Diffusers isc as W

MAINE YANKEE ATOMIC POWER CO. Site Characterization Locations Of Radiological Figure LICENSE TERMINATION PLAN Survey Packages and Elevated Areas 2-1

MYAPC License Termination Plan Revision 3 October 15, 2002 1

RO W UT ES TP E OR AD WISCASSET T E RO U.S. ID G

BR N RO 144 UT E U TE MAINE 14 4 YANKEE RO NON-IMPACTED NON-IMPACTED RI VE R

BA CK 14 4

IMPACTED RO UT LITTLE OAK E

ISLAND BAILEY POINT FOXBIRD ISLAND WESTPORT ISLAND MONTSWEAG BAY MAINE YANKEE ATOMIC POWER CO. Site Characterization Figure LICENSE TERMINATION PLAN Impacted / Non- Impacted Areas 2-2

MYAPC License Termination Plan Revision 3 October 15, 2002 ATTACHMENT 2E Site and Survey Area Maps

MYAPC License Termination Plan Revision 3 October 15, 2002 N

115KV SWITCHYARD 13 14 15 16 17 18 27 30 33 36 39 41 43 45 46 48 49 53 55 58 64 67 68 71 73 74 104 81 103 86 83 REACTOR 87 89 90 BUILDING 98 96 102 100 SERVICE BUILDING TURBINE HALL XX Approximate Survey Location MAINE YANKEE ATOMIC POWER CO. Site Characterization West Side Yard Figure LICENSE TERMINATION PLAN Survey Package R0100 2-3

MYAPC License Termination Plan Revision 3 October 15, 2002 2

5 6 REACTOR 16 BUILDING 21 SERVICE 38 BUILDING 42 44 46 55 TURBINE HALL 68 75 77 78 86 102 109 114 132 143 152 155 158 159 161 151 212 214 215 222 214 Approximate Survey Location MAINE YANKEE ATOMIC POWER CO.

Site Characterization East Side Yard Figure LICENSE TERMINATION PLAN Survey Package R0200 2-4

MYAPC License Termination Plan Revision 3 October 15, 2002 "6M" FIRE POND "6L" N "5C" PARKING "6K" "6J" WAREHOUSE ADMINISTRATION "5B" BUILDING "6D" "6I" "6C" "6E" 115KV SWITCHYARD "6G" "6F" "5A" B-206 "6H" 005 BK-1 "6B" "7 F" "7 G" "7H" "6 A" REACTOR BUILDING "7 C" I CE B 006 TURBINE HALL "7 E" UILD ING B-203 S ER V "7B" B-206 RD "7D" B-205 B-202 B-201 007 MAINE YANKEE ATOMIC POWER CO. Site Characterization Roof and Yard Drains Figure LICENSE TERMINATION PLAN #005, 006, 007; Survey Package R0300-1 2-5

MYAPC License Termination Plan Revision 3 October 15, 2002 ING 30 PAR K N13 012 ADMI N

"11E" "11C" "11A" "11D" "11B" "10F" "10H" "10D" 011 "10E" "10C" "10A" SP-12 B U IL D IN E H 010 ALL ING "10B" MW-1OO TURB ICE SP-6 SER V "9A" "9E" "9D" 009 SP-7 Post Indicator "9B" Valve "9C" 008 MAINE YANKEE ATOMIC POWER CO. Site Characterization Roof and Yard Drains Figure LICENSE TERMINATION PLAN 2-6

MYAPC License Termination Plan Revision 3 October 15, 2002 BAILEY COVE N

5 4

3 1

FOXBIRD 2 ISLAND 20 10 115KV SWITCHYARD FOREBAY 20 0 6 7

RE A C BUILD TOR 9 8 ING 10 MONTSWEAG SERV BAY ICE BUILD ING 0 TU R B I NE H ALL 30 BAILEY POINT 20 10 BAC K RI VER X

MAINE YANKEE ATOMIC POWER CO. Site Characterization Forebay Shorelines Figure LICENSE TERMINATION PLAN Survey Package R0400-1A 2-7

MYAPC License Termination Plan Revision 3 October 15, 2002 0 20 10 MONTSWEAG BAY 0

2 5

7 12 9 18 14 19 15 21 17 BAILEY 27 POINT 25 36 45 34 47 49 42 56 57 59 53 60 54 62 63 64 65 66 10 XX - APPROXIMATE SAMPLE LOCATION MAINE YANKEE ATOMIC POWER CO. Site Characterization Bailey Point Figure LICENSE TERMINATION PLAN Survey Package R0500-2 2-8

MYAPC License Termination Plan Revision 3 October 15, 2002 0 20 10 MONTSWEAG BAY 0

This Figure Deleted (duplicate of Figure 2-8) 5 7

2 12 9 18 14 19 15 21 17 BAILEY 27 POINT 25 36 45 34 47 49 42 56 57 59 53 60 54 62 63 64 65 66 10 XX - APPROXIMATE SAMPLE LOCATION MAINE YANKEE ATOMIC POWER CO. Site Characterization Bailey Point Figure LICENSE TERMINATION PLAN Survey Package R0500-2 2-9

MYAPC License Termination Plan Revision 3 October 15, 2002 0 20 10 1

2 MONTSWEAG 5 3 BAY 4 6 7 8

11 10 12 0 13 14 17 15 18 16 19 20 21 22 23 24 25 BAILEY 26 POINT 27 28 30 31 32 33 10 35 36 XX - APPROXIMATE SAMPLE LOCATION MAINE YANKEE ATOMIC POWER CO. Site Characterization Bailey Point Figure LICENSE TERMINATION PLAN Survey Package R0500-4 2-10

MYAPC License Termination Plan Revision 3 October 15, 2002 30 20 10 AD RO 60 Y

RR FE D

OL 30 MAINE YANKEE ATOMIC POWER CO. Site Characterization Ball Field Figure LICENSE TERMINATION PLAN Survey Package R0600-1 2-11

MYAPC License Termination Plan Revision 3 October 15, 2002 30 4

14 20 10 24 34 44 54 1 64 2 3

74 13 23 83 33 43 91 53 63 73 82 RO 98 90 AD 97 60 101 Y

RR FE D

OL 30 MAINE YANKEE ATOMIC POWER CO. Site Characterization Ball Field Figure LICENSE TERMINATION PLAN Survey Package R0600-2 2-12

MYAPC License Termination Plan Revision 3 October 15, 2002 N 30 4

14 20 10 24 15 34 25 44 26 54 36 27 1 19 64 2 12 3 30 74 66 13 49 23 75 59 50 33 83 51 43 Biased 61 91 86 79 71 62 53 Survey 81 72 63 91 73 82 RO 98 96 90 AD 97 60 101 FE RR Y

OL D

30 XX Approximate Survey Location MAINE YANKEE ATOMIC POWER CO. Site Characterization Ball Field Figure LICENSE TERMINATION PLAN Survey Package R0600-2 2-13

MYAPC License Termination Plan Revision 3 October 15, 2002 0

0 1 2 20 7 10 3 10 4

20 14 15 30 25 BALL 26 27 19 FIELD 45 47 36 37 61 51 40 64 52 76 67 56 78 90 81 92 73 94 84 95 T EM 99 GEN PORARY ENC ERATOR 100 102 L OS URE RO 30 AD RR

1. 1. Approximate Survey locations Y

1. 1. Reference Points FE 40 D OL MAINE YANKEE ATOMIC POWER CO. Figure LICENSE TERMINATION PLAN Survey Package R0700-2 2-14

MYAPC License Termination Plan Revision 3 October 15, 2002 Bailey Cove W A RE HOUS 115K V E SWIT CHYA RD FIRE POND RE A C 1 BUILD TOR 2 ING 4

6 345KV SWI T CHYA 12 RD SERV 13 BUILDICE ING 19 ADM BUILD IN ING TURB INE H ALL 31 PARK ING 33 34 25 35 26 45 36 29 59 60 41 52 43 73 44 L.L.W 75 STOR .

AGE 87 67 68 89 81 70 90 82 101 72 114 116 117 108 109 P A RK 119 110 ING 127 122 113 128 130 125 136 B A CK 137 RIVER LITTLE OAK ISLAND

1. 1. 1. Approximate Survey Location MAINE YANKEE ATOMIC POWER CO. Site Characterization Admin. and Parking Areas Figure LICENSE TERMINATION PLAN Survey Package R0800-2 2-15

MYAPC License Termination Plan Revision 3 October 15, 2002 BAILEY COVE 39 40 80 98 83 115KV WARE HOUS 48 100 E 50 101 SWIT CHYA 69 RD FIRE FO POND 52 RE BA Y

109 93 73 130 108 REAC 120 107 BUILDTOR 78 ING 123 94 95 111 345KV 113 SWIT 115 SERV CHYA ICE RD BUILD 126 ING ADMIN 128 T URB INE H ALL PARK ING 151 153 L.L.W STO .

RA GE 158 TEM GEN PORARY ENC ERATOR L OS URE PA RK ING PL AN T

AC LITTLE BA CE OAK CK SS RI ISLAND VE RO R

AD XX Approximate Sample Locations MAINE YANKEE ATOMIC POWER CO. Site Characterization Balance of Plant Areas Figure LICENSE TERMINATION PLAN Survey Package R0900 2-16

MYAPC License Termination Plan Revision 3 October 15, 2002 BAILEY COVE 5

12 14 17 19 FOXBIRD 34 39 ISLAND 44 54 63 65 68 69 71 73 76 78 81 83 85 FOREBAY 86 88 91 93 94 98 102 118 122 125 MONTSWEAG BAY BAILEY POINT XX Approximate Survey Locations MAINE YANKEE ATOMIC POWER CO. Site Characterization Foxbird Island Figure LICENSE TERMINATION PLAN Survey Package R1000 2-17

MYAPC License Termination Plan Revision 3 October 15, 2002 ING 30 PAR K N13 012 ADMI N

"11E" "11C" "11A" "11D" "11B" "10F" "10H" "10D" 011 "10E" "10C" "10A" SP-12 B U IL D IN E H 010 ALL ING "10B" MW-1OO TURB ICE SP-6 SER V "9A" "9E" "9D" 009 SP-7 Post Indicator "9B" Valve "9C" MAINE YANKEE ATOMIC POWER CO. Site Characterization Roof and Yard Drains Figure LICENSE TERMINATION PLAN #005, 009-12; Survey Package R1100-1 2-18

MYAPC License Termination Plan Revision 3 October 15, 2002 005 "5A" 006 "5B" 115KV WAREHOUSE "5C" SWITCHYARD "6A" "7F" B-203 SP-9 "6B" 007 SP-10 "6C" "7E" "7G" BK-1 "7D" "6J" "7I" "7H" B-202 SP-5 RE A C "6F" "6E" "7A" B UI L D T O R "6D" "6K" ING B-206 "6L" "7B" B-201 SERV "7C" "6G" "6I" ICE B UILDI NG B-204 "6H" ADMI N

T URB I NE H ALL PARK ING INFOR MATIO N CEN TER MAINE YANKEE ATOMIC POWER CO. Site Characterization Roof and Yard Drains Figure LICENSE TERMINATION PLAN 2-19

MYAPC License Termination Plan Revision 3 October 15, 2002 Outfall #017 20 30 60 L.L STOR.W.

AGE TEM POR GEN ARY ERA ENC TOR LO S URE PA 40 RK IN G

40 PL AN T

AC CE SS RO AD 30 MAINE YANKEE ATOMIC POWER CO. Site Characterization Roof and Yard Drains Figure LICENSE TERMINATION PLAN #017 Survey Package R1100-1 2-20

MYAPC License Termination Plan Revision 3 October 15, 2002 2

3 8 6 9 7 10 11 15 16 13 17 14 18 22 20 L.L.W ST OR .

25 AGE 23 26 24 27 28 29 32 30 33 31 36 38 39 XX Approximate Survey Locations MAINE YANKEE ATOMIC POWER CO. Figure LICENSE TERMINATION PLAN 2-21

MYAPC License Termination Plan Revision 3 October 15, 2002 20 30 60 L.L STOR.W.

AGE TEM POR 46 GEN ARY 22 ERA 11 ENC TOR LO S URE 15 27 88 28 102 39 53 103 79 40 55 PARKING 68 145 107 95 134 72 110 73 137 113 136 137 138 116 142 156 40 PL AN T

AC CE SS RO AD 30 XX Approximate Survey Locations MAINE YANKEE ATOMIC POWER CO. Site Characterization Dry Cask Storage Area Figure LICENSE TERMINATION PLAN Survey Package R1300-2 2-22

MYAPC License Termination Plan Revision 3 October 15, 2002 N

RIV ER PL AN T CK 12 INTAKE BA 13 6

DISCHARGE LITTLE FOREBAY OAK ISLAND BAILEY 14 POINT 7

8 1

FOXBIRD 15 ISLAND 9

11 2 10 ISL AN D

3 RT LONG EDG PO ST E

WE 4

5 MONTSWEAG BAY DISCHARGE AREA (DIFFUSER)

XX APPROXIMATE SURVEY LOCATION MAINE YANKEE ATOMIC POWER CO. Figure LICENSE TERMINATION PLAN 2-23

MYAPC License Termination Plan Revision 3 October 15, 2002 N

I 35 G 21 34 F 14 27 7 20 33 E 13 26 6 19 D 12 25 5 18 C 11 24 17 30 B 10 23 3 16 29 A 9 22 2 15 AA 8 BB 1 CC DD EE FF ASH ROAD XX Approximate Survey Location MAINE YANKEE ATOMIC POWER CO. Site Characterization Ash Road Rubble Piles Figure LICENSE TERMINATION PLAN Survey Package R1500-2 2-24

MYAPC License Termination Plan Revision 3 October 15, 2002 RO 49 AD 54 N RE AD PO Y

INT 46 YOUNG BROOK 51 45 56 61 RO 50 AD FE RR Y

OL D

625,000 E 410,000 N 73 AD RO 70 AC TEMP CE GENE ORARY SS 69 ENCL RATOR OSUR E

PL AN T

PARKING L.L.W 67 STOR .

AGE 345KV SWITC RI HYARD VE R

FIRE BA 63 POND CK PARKING WAR HOUSE-E ADMIN 115KV SWITCHYARD TOR SERV ICE TURB INE HA REAC BUILD ING LL LITTLE OAK ISLAND FOREBAY CHEWONKI CREEK 24 41 5 14 23 33 40 BAILEY BAILEY COVE POINT 39 YOUNG POINT ROAD 31 11 30 1 29 FOXBIRD 28 ISLAND 8 17 27 MONTSWEAG BAY 620,000 E 405,000 N YOUNG POINT XX Approximate Survey Locations MAINE YANKEE ATOMIC POWER CO. Site Characterization Owner Controlled Area Figure LICENSE TERMINATION PLAN Survey Package R1600-4 2-25

MAINE YANKEE MYAPC License Termination Plan ATOMIC POWER CO. A B C D E F G H I J K L M N O P Q R S T U V W X Y Z YOUNG 620,000 E 620,000 E POINT CHEWONKI CREEK N LICENSE 405,000 N 410,000 N Revision 3 TERMINATION PLAN AA BB October 15, 2002 CC YOUNG POINT ROAD DD EE FF RE AD Y PO INT GG RO AD HH Site Characterization Owner Controlled Area II BAILEY COVE JJ KK FOXBIRD WAR 115KV HOUSE-SWITCHYARD E FIRE ISLAND POND FOREBAY REAC TOR 345KV SERV SWITC ICE HYARD BUILD ING ADMIN MONTSWEAG BAY TURB INE HA LL PARKING BAILEY L.L.W STOR . RO AGE POINT TEMP GENE ORARY AD ENCL RATOR OSUR E

FE RR PARKING Y OL D

PL AN Survey Package R1600-4 T

AC CE 625,000 E SS LITTLE BA RO OAK CK AD 410,000 N ISLAND RI VE R

Figure Drive-over Area Scanned 2-26 Elevated Reading

MYAPC License Termination Plan Revision 3 October 15, 2002 N

75 43 73 41 71 39 38 38 15 35 63 32 62 31 61 60 59 28 57 7 56 25 55 54 23 4 22 52 21 Approximate Survey Location H R OA D

AS 18 YOUNG BROOK 45 OL XX D

FE RR Y

RO AD RI VE R

RO CK AD SS BA CE AC MAINE YANKEE ATOMIC POWER CO. Site Characterization Owner Controlled Area Figure LICENSE TERMINATION PLAN North of Old Ferry RD Survey Package R1700-1 2-27

MYAPC License Termination Plan Revision 3 October 15, 2002 N

75 92 25 72 51 87 121 12 23 103 120 158 343 83 133 144 132 9 20 45 64 H R 141 OA D

AS 18 60 YOUNG BROOK 17 42 78 96 OL D

FE RR Y

RO AD 76 RI VE R

RO CK AD SS BA XX Approximate Survey Location CE AC MAINE YANKEE ATOMIC POWER CO. Site Characterization Owner Controlled Area Figure LICENSE TERMINATION PLAN North of Old Ferry RD Survey Package R1700-4 2-28

MYAPC License Termination Plan Revision 3 October 15, 2002 N

YOUNG BROOK FIRE POND 345KV SWITC HY ARD L.L.W RO STO .

RA GE AD TEM P

GEN ORARY ENC ERATOR LOS URE RR Y

FE PARKING D

OL PL AN T

AC CE SS RO AD BA CK RI VE R

AS H

RO AD MAINE YANKEE ATOMIC POWER CO. Site Characterization Bailey House Area Figure LICENSE TERMINATION PLAN Survey Package R1800-1 2-29

MYAPC License Termination Plan Revision 3 October 15, 2002 2 1 3 4 N

RIV ER PL AN T CK INTAKE BA 17 20 FOREBAY LITTLE OAK ISLAND BAILEY 23 POINT FOXBIRD ISLAND ISL AN D

RT LONG EDG PO ST E

WE MONTSWEAG BAY DISCHARGE AREA (DIFFUSER)

XX APPROXIMATE SURVEY LOCATION MAINE YANKEE ATOMIC POWER CO. Site Characterization Bailey Cove Figure LICENSE TERMINATION PLAN Survey Package R1900 2-30

MYAPC License Termination Plan Revision 3 October 15, 2002 N

RIV ER PL AN T CK INTAKE BA FOREBAY LITTLE OAK ISLAND BAILEY POINT FOXBIRD ISLAND ISL 1 AN D

2 RT 3

LONG EDG PO ST E

4 WE 5

DISCHARGE AREA (DIFFUSER)

MONTSWEAG BAY XX APPROXIMATE SURVEY LOCATION MAINE YANKEE ATOMIC POWER CO. Site Characterization Diffusers Figure LICENSE TERMINATION PLAN Survey Package R2000 2-31

MYAPC License Termination Plan Revision 3 October 15, 2002 TEM P

GEN ORARY ENC ERATOR LOSU RE 1

2 3

4 7 5 8 6 9

10 13 11 14 12 15 16 19 17 20 18 21 22 25 23 26 24 27 28 29 30 XXX Approximate location of sample and survey point MAINE YANKEE ATOMIC POWER CO. Figure LICENSE TERMINATION PLAN Survey Package R2100 2-32

MYAPC License Termination Plan Revision 3 October 15, 2002 N

Grid 1 Grid 2 28/28 32/26 27/29 28/29 27/28 1 2 28/30 24/28 25/26 3 4 1 2 PWST 3 4 BWST A BWST B

Fuel P. A. B.

Building BUILD ING I CE S E RV REACTOR BUILDING General Area Survey Results 6/36 uR/hr Ludlum 2350-1

1. 126182 3-22-98 MAINE YANKEE ATOMIC POWER CO. Figure LICENSE TERMINATION PLAN Survey Package R2300-1 2-33

MYAPC License Termination Plan Revision 3 October 15, 2002 N

1 2

Scan #1 Grid ID 3

Scan #2 4 Grid ID 1

2 3

4 Transf ormer X16 BWST B

Transf ormer X14 P. A. B.

= Area Scan X = Sample Point Point location MAINE YANKEE ATOMIC POWER CO. Figure LICENSE TERMINATION PLAN Survey Package R2300-2 2-34

MYAPC License Termination Plan Revision 3 October 15, 2002 0 20 10 MONTSWEAG BAY 0

BAILEY POINT 10

- APPROXIMATE LOCATION OF ELEVATED AREA MAINE YANKEE ATOMIC POWER CO. Site Characterization Drive Over Elevated Areas Figure LICENSE TERMINATION PLAN Survey Package R2500-1 2-35

MYAPC License Termination Plan Revision 3 October 15, 2002 30 20 10 RO AD 60 FE RR Y

OL D

30 Approximate location of elevated area MAINE YANKEE ATOMIC POWER CO. Site Characterization Drive Over Area Figure LICENSE TERMINATION PLAN Survey Package R2500-2 2-36

MYAPC License Termination Plan Revision 3 October 15, 2002 0

0 10 20 6

7 BALL 8 FIELD 5

4 1 3 TEM GEN PORARY 2

ENC ERATOR LOS URE RO 30 AD RR Y

FE Approximate location of elevated area 40 OL D

MAINE YANKEE ATOMIC POWER CO. Figure LICENSE TERMINATION PLAN Survey Package R2500-3 2-37

MYAPC License Termination Plan Revision 3 October 15, 2002 Bailey Cove WARE HOUS 115KV E SWIT CHYA RD FIRE POND REA BUILC TO DINGR 345KV SWIT CHYA RD SERV BUILDICE ING ADM BUILD IN ING TURB INE H PARK ALL ING L .L .W STOR .

AGE 1

PARK ING 2

3 BACK RIVER LITTLE OAK ISLAND Approximate location of elevated area MAINE YANKEE ATOMIC POWER CO. Site Characterization Drive Over Areas Figure LICENSE TERMINATION PLAN Survey Package R2500-4 2-38

MYAPC License Termination Plan Revision 3 October 15, 2002 20 30 60 L.L STOR.W.

AGE TEM POR GEN ARY ERA E NC TOR LO S URE PA 40 RK IN G

2 1

40 PL AN T

AC CE SS RO AD 30 Approximate location of elevated areas MAINE YANKEE ATOMIC POWER CO. Site Characterization Drive Over Elevated Areas Figure LICENSE TERMINATION PLAN 2-39

MAINE YANKEE MYAPC License Termination Plan A B C D E F G H I J K L M N O P Q R S T U V W X Y Z ATOMIC POWER CO.

YOUNG 620,000 E 620,000 E POINT CHEWONKI CREEK LICENSE 405,000 N 410,000 N N Revision 3 TERMINATION PLAN AA BB October 15, 2002 CC YOUNG POINT ROAD DD EE FF RE AD Y PO GG INT RO AD HH II BAILEY COVE Site Characterization Drive Over Elevated Areas Figure JJ YOUNG BROOK KK FOXBIRD WAR ISLAND 115KV HOUSE-SWITCHYARD E FIRE FOREBAY POND REAC TOR 345KV SERV SWITC ICE HYARD BUILD MONTSWEAG BAY ING ADMIN TURB INE HA LL PARKING BAILEY L.L.W RO POINT STOR .

AGE AD TEMP GENE ORARY ENCL RATOR OSUR E RR Y

FE PARKING D

OL PL AN T

AC CE 625,000 E SS LITTLE OAK BA RO AD ISLAND CK 410,000 N RI VE R

Survey Package R2500-7 2-40 Approximate location of elevated area

MYAPC License Termination Plan Revision 3 October 15, 2002 N

YOUNG BROOK FIRE POND 345KV SWITC HY ARD L.L.W RO STO .

RA GE AD TEM P

GEN ORARY ENC ERATOR LOS URE RR Y

FE PARKING D

OL 1

PL AN T

AC 2 3 CE SS RO 4 AD BA CK RI VE R

AS H

RO AD Approximate location of elevated areas MAINE YANKEE ATOMIC POWER CO. Site Characterization Drive Over Elevated Areas Figure LICENSE TERMINATION PLAN Survey Package R2500-8 2-41

MYAPC License Termination Plan Revision 3 October 15, 2002 N

32 33 34 35 36 137 26 < 2 pci/gm CS 31 25 30 24 29 23 28 FOREBAY 22 20 27 21 18 19 12 17 6 11 16 5 10 15 4 9

14 3 8

13 2 1 7 6 1 R2501 / 01OA1 5 RWST 10 9 19 20 / 01OB1 14 18 R2501 / 02OA1 2 13 17 3

26 2 8 7

4 5 2 4 12 23 2 2 21 11 16 1 5

XX -Approximate Survey Location MAINE YANKEE ATOMIC POWER CO. Site Characterization Forebay Area Figure LICENSE TERMINATION PLAN Survey Package R2501 2-42

MYAPC License Termination Plan Revision 3 October 15, 2002 L.L.W STOR .

AGE T e mp Gene orary Enclo rator sure Maint ena Stora nce ge Yard PARKING 2 8 6 22 3 1 R2501/03OA1 7

25 5 11 /03OA1 9

19 4 1 10 23 18 3 14 21 12 Pl 16 24 an 15 20 t Ac 17 ce ss Ro ad XX- Approximate Survey Locations MAINE YANKEE ATOMIC POWER CO. Site Characterization Dry Cask Storage Area Figure LICENSE TERMINATION PLAN Survey Package R2501 2-43

MYAPC License Termination Plan Revision 3 October 15, 2002 N

7 6 23 9 8 22 15 14 2 1 1 11 10 3

17 16 5 13 12 2 4 25 19 18 24 21 20 Dry Cask Storage Area X 0 - 6" Soil Sample Location (Approximate)

X 6 - 12" Soil Sample Location (Approximate)

MAINE YANKEE ATOMIC POWER CO. Site Characterization Follow-up Sampling At Elevated Figure LICENSE TERMINATION PLAN Soil Sample Locations Survey Package 2-44

MYAPC License Termination Plan Revision 3 October 15, 2002 5

15 1 4 3

14 15 2 1 14 10 9

2 13 1 SIT#1 SIT#3 1 13 14 3

13 2 12 RCP#1 1 2

8 S/G#3 3

12 4

S/G#1 3

12 2

4 RCP#3 11 11 4 5 11 IC SUMI P 2 REACTOR P 10 HEAD QU ZR 1 EN LAYDOWN TAN CH K 5 5 3 5 4

3 9

RCP#2 8 SIT#2 S/G#2 6 4 10 7 6 5

7 6 6 10 8 9 7 Stairway 01SW2 7 9

8 Stairway 01SW1 Floor 01FL1 Walls 01WS2 Walls 01WS1 Equipment 01EQ1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Containment Building -2ft Elevation Figure LICENSE TERMINATION PLAN Survey Package A0100 2-45

MYAPC License Termination Plan Revision 3 October 15, 2002 1

20 1

30 2 2

19 20 CEA CHANGE STA 1 3

1 19 5 FUEL TRANSFER 1 3 4 CANAL 2 18 29 3 18 17 UPPER GUIDE 2 STRUCTURE 4 SIT#3 STORAGE 16 AREA 10 SIT#1 5

28 9 4 3

17 15 CORE BARREL& 6 S/G#3 THERMAL SHIELD RCP#1 2 STORAGE 14 AREA 27 5 7

26 S/G#1 3 4 16 8

13 RCP#3 4

8 25 5 12 PZR 1 7 9 6 5 5 6 15 24 4

2 3 23 22 10 14 RCP#2 SIT#2 S/G#2 7 11 21 20 7 12 11 10 9 8 15 13 14 19 13 6

8 12 18 16 9 11 17 10 XX NW Stair well 01SW2 XX SW Stairwell 01SW1 Equipment 01EQ1 XX XX Outer Walls 01WS1 XX Inner Walls 01WS2 XX Floors 01FL1 MAINE YANKEE ATOMIC POWER CO. Site Characterization 20 ft Elevation Figure LICENSE TERMINATION PLAN Survey Package A0200 2-46

MYAPC License Termination Plan Revision 3 October 15, 2002 10 11 13 14 13 7 10 15 9 8 8

7 16 90 14 11 6 2 15 9 12 1 8 6 18 17 12 10 5

9 16 4 19 18 3 4 5 3 5 2 5 17 1

1 20 3 2 91 P-17B 2 2 4 6 4

3 1 This area not 3

P-17A surveyed due to 2

high dose 7 rates.

1 1 92 P-85 This area not surveyed due to high dose rates.

XX Stair well XX Floor Drains Equipment 01EQ1 XX XX Walls XX Floor Drain XX Floors MAINE YANKEE ATOMIC POWER CO. Site Characterization Fuel Building 21 ft Elevation Figure LICENSE TERMINATION PLAN Survey Package A0400 2-47

MYAPC License Termination Plan Revision 3 October 15, 2002 N

Spray Building 6 6 5 4

3 6 5 4 3 3 3 7

5 2 1

5 2

2 6

2 1

8 2 1 1 1

4 3

3 4

2 9 1 2 1

XX Tank Exterior 01WE1 XX Wall Exterior 02WE1 Ceiling 02CL1 XX XX Wall Interior 02WS1 XX Equipment 02EQ1 XX Structural Tank Supports 01SS1 XX Floor 02FL1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Demineralized Water Figure LICENSE TERMINATION PLAN Storage Tank (DWST) Survey Package A0500 2-48

MYAPC License Termination Plan Revision 3 October 15, 2002 FL-8A FL-8B I-4 I-5A I-5B R-45 I-2C I-2B I-2A I-3 2 3 P-24A P-24B 1 4 P-18A P-18B 2

2 1 3 5 5 4 3 2 1 3

8 6 7 4 1 6 Letdown HXCHs P-11 P-22 FL-35A FL-35B FL-34A FL-34B P-20 P-19 30 19 18 17 29 20 P-8 28 27 7

P-65 P-21 26 25 TK-3 6 5 P-81 FL-33B 23 8

22 FL-33A 24 16 P-66A P-66B 21 E-92A/B 4

20 P-7 16 15 DA-1A DA-1B 3 17 15 19 14 18 13 14 12 P-16A P-16B P-23B P-23A 13 2 1 11 12 P-15A P-15B 11 10 TK-12A TK-12B 10 TK-11 9 9

8 P-67A/B XX Equipment Ceiling XX XX Floors XX Walls MAINE YANKEE ATOMIC POWER CO. Site Characterization Primary Auxiliary Building 11 ft Figure LICENSE TERMINATION PLAN Survey Package A0600 2-49

MYAPC License Termination Plan Revision 3 October 15, 2002 7 6 6 6 5 Survey Unit 03 5

2 3 1 4 4 5 Chemistry Lab 4

Survey Unit 02 1 2 3 3 1 2 3 Chemistry 3 1

2 Offices 1 2 2 Meter 3 8 6 4 Cal 5

RP 4 2

9 Count Room 4 7 14 9 5 5 1 8 13 6 RP 5

Meter Checkpoint 1 1 Room 11 3 15 2 10 12 7 4 RCA 2 RCA 3

6 Boundary 2 Boundary 4 2 3

3 1

1 1 1 1 Women's 6

Restroom 1

RP Survey Unit 01 Dress Out Offices Room Survey Unit 04 7 7

Dosimetry 1

4 2 Men's 2 Restroom 8 2 3 4

Respirator 5 Room 4

8 3 9

8 10 9 Men's 6 Locker Room Survey Unit 05 1

1 1

05EQ3 05EQ1 1

1 05EQ2 7

2 7 Hot Shop 5

3 1

05EQ6 1 1 05EQ4 05EQ5 4 4 1

6 1 3 1

Tool 1

1 2 Room 5

11 8 MOVAT Room Survey Unit 06 2 1 Survey Unit 07 XX Equipment 1 Survey Unit 08 Seal Rebuild Room Ceiling XX 1 2

XX Floors XX Walls MAINE YANKEE ATOMIC POWER CO. Site Characterization Service Building Hot Side 21 ft Figure LICENSE TERMINATION PLAN Survey Package A0900 2-50

MYAPC License Termination Plan Revision 3 October 15, 2002 2 3 4

1 2 3 N 1 4

5 6

5 6 6

9 7

44 45 43 10 7 8 8

42 9 11 13 41 12 11 10 12 14 1

13 14 15 16 40 17 15 23 18 5 39 16 24 17 38 19 25 18 37 2 20 19 36 21 26 20 35 21 22 3 4 29 28 22 27 33 34 23 32 30 30 Walls Lower (01WL1) 29 28 27 26 25 24 XX XX Equipment (01EQ1)

XX 31 Floors (01FL1)

MAINE YANKEE ATOMIC POWER CO. Site Characterization Low Level Waste Storage Building 21 ft Figure LICENSE TERMINATION PLAN Survey Package A1100 2-51

MYAPC License Termination Plan Revision 3 October 15, 2002 2 1 3 2 1 24 6

N 5 4 3 2 1

23 3

1 2

22 4 6 4

5 8 9 10 11 12 7

6 21 7

20 18 17 16 15 14 13 8

9 7 8 19 9

18 19 20 11 23 12 24 10 21 22 10 17 11 16 13 28 27 26 25 30 15 29 12 14 15 13 17 36 31 16 32 33 34 35 18 14 4

XX Stairs (01SW1) 3 Walls Upper (01WU1)

XX XX Ceiling #2 (01CL2)

XX Ceiling #1 (01CL1)

MAINE YANKEE ATOMIC POWER CO. Site Characterization Low Level Waste Storage Building 21 ft Figure LICENSE TERMINATION PLAN Survey Package A1100 2-52

MYAPC License Termination Plan Revision 3 October 15, 2002 N

1 2 3 4 5 6 5 1 2 5

14 5 7 6 1

1 4

13 3

7 2 4 3 2 12 11 3 10 9 4 8 XX Equipment (02EQ1)

XX Walls (01WS1)

XX Ceiling #2 (02CL1)

XX Floor (02FL1)

MAINE YANKEE ATOMIC POWER CO. Site Characterization Low Level Waste Storage Building 21 ft Figure LICENSE TERMINATION PLAN Office Survey Package A1100 2-53

MYAPC License Termination Plan Revision 3 October 15, 2002 N 1 2 3 4 1 2 3 5

5 1

9 4 6 8 XX 7

2 3 1 9

7 10 4 12 5 2 6 8 4 8 10 11 6 6 3

5 7

6 8 12 TK-95 9

7 10 11 5

13 TK-85 14 4

Berm Berm 3 17 Decon Pad 15 16 18 Jib HIC Crane Jib 2 Crane 1 19 10 9

TK-109 20 XX Lower Walls (01WL1) 8 XX Upper Walls (01WU1)

Equipment (01EQ1)

XX XX Ceiling (01CL1)

XX Floor (01FL1)

MAINE YANKEE ATOMIC POWER CO. Site Characterization RCA Storage Building Figure LICENSE TERMINATION PLAN Survey Package A1200 2-54

MYAPC License Termination Plan Revision 3 October 15, 2002 N

6 9

7 10 8 8

3 9

3 12 11 5

2 7

10 5 3 4

1 3 1 2

2 2 1 1 4 4

5 6

XX Walls (01WS1)

XX Equipment (01EQ1)

XX Ceiling (01CL1)

XX Floor (01FL1)

MAINE YANKEE ATOMIC POWER CO. Site Characterization Equipment Hatch Area 21ft Figure LICENSE TERMINATION PLAN Survey Package A1300 2-55

MYAPC License Termination Plan Revision 3 October 15, 2002 N

Wall Vent 11 10 1

1 12 2

6 Personnel 3 Hatch 2

3 4 1 2 01FL1 130 3

01CL1 2 01EQ1 7

7 01WS1 10 3 6 1 5 4 5

9 6

8 5 9 4 4 5 8

UP 2 5 1 4 02FL1 2

2 02CL1 4

4 02EQ1 Ventilation 02WS1 3

Units 129 1

1 3

2 3 Filter Banks XX Walls XX Equipment XX Ceiling XX Floor MAINE YANKEE ATOMIC POWER CO. Site Characterization Personnel Hatch Area 21ft Figure LICENSE TERMINATION PLAN Survey Package A1400 2-56

MYAPC License Termination Plan Revision 3 October 15, 2002 N

1 2 1

Mechanical Penetration Room TK-25 4 3 2

01EQ2 2 1 4

3 2

P-25B 5

3 4

1 6 8

7 1

8 2 01EQ1 7

9 10 6 5 XX Walls XX Equipment XX Floor Drain XX Floor MAINE YANKEE ATOMIC POWER CO. Site Characterization Mechanical Penetration Room 21 ft Figure LICENSE TERMINATION PLAN Survey Package A1500 2-57

MYAPC License Termination Plan Revision 3 October 15, 2002 N 2 1 2

1 1

3 2 1 3 1 2 4

1 1 1 5 6 2

1 3

4 1 1 1 4 3 2 1 XX Equipment XX Ceiling XX Floor MAINE YANKEE ATOMIC POWER CO. Site Characterization Mechanical Penetration Room Figure LICENSE TERMINATION PLAN Elevations 2, 3, 4, and 5 Survey Package A1500 2-58

MYAPC License Termination Plan Revision 3 October 15, 2002 N

3 2

2 bk bk 1

bk 3

1 bk bk 2

bk Background bk XX Equipment 1 1

XX Floor XX Walls MAINE YANKEE ATOMIC POWER CO. Site Characterization Electrical Penetration Room Figure LICENSE TERMINATION PLAN Bottom Level Survey Package A1600 2-59

MYAPC License Termination Plan Revision 3 October 15, 2002 N

Spray Building Elevation 4'-0 14 13 2 7 1 11 4 5

4 5

9 3 12 3 3 4 10 4 5 6 6 7 3 8 2 8 2

2 1 6

1 1

7 18 1

2 Spray Building Elevation 6-0 10 17 6 4 2 8 8 15 2 11 4 19 6 3 14 20 7

1 1 9 16 2 3 12 5 7

1 10 9 13 5 3 4

XX Walls XX Equipment XX Ceiling XX Floor XX Stairs MAINE YANKEE ATOMIC POWER CO. Site Characterization Containment Spray Building Figure LICENSE TERMINATION PLAN 4 ft and 6 ft Elevations Survey Package A1700 2-60

MYAPC License Termination Plan Revision 3 October 15, 2002 02FL1- Floor 02WS1- Walls N 02EQ1- Equipment 02CL1- Ceiling 7 6 5 4 3 5

2 1

3 5 4 3 1 8

2 4

9 7 2 5 6 4 7 UP 10 6 1 3 2 1 UP DN 12 02SW2 Elevation 21'-0" 02SW1 2 samples 2 samples 03- Survey Unit 2 1 2 4

1 3 UP 4 3 2 1 Elevation 12'-0" XX Walls 03WS1 XX Ceiling 03CL1 XX Floor 03FL1 XX Stairs 03SW1 Floor Opening MAINE YANKEE ATOMIC POWER CO. Site Characterization Containment Spray Building Figure LICENSE TERMINATION PLAN 12 ft and 21 ft Elevations Survey Package A1700 2-61

MYAPC License Termination Plan Revision 3 October 15. 2002 N

2 3 1 10 7 2 10 8 5 6 4 1 7 6 9

4 8

3 9 5 Spray Building 14'-0" (Motors)

XX Walls XX Floor MAINE YANKEE ATOMIC POWER CO. Site Characterization Containment Spray Building Figure LICENSE TERMINATION PLAN 14 ft Elevation Survey Package A1700 2-62

MYAPC License Termination Plan Revision 3 October 15, 2002 N

Rock CTMT Room APD P-25-A 11 7 ALT 8 S.D. 2 Panel 9

6 1 1 2

4 1 12 6 5 2 10 7

3 10 8 TK-89 4 Chemical 5 1 Addition 3 Tank 9

CTMT 5 Purge 6 Valves 4

AFW-A-39 3 ATW-A-201 3 P-25-C 4

AFW-A-101 2 Primary Vent Stack XX Walls 01WS1 XX Floor 01Fl1 XX Equipment 01EQ1 XX Ceiling 01CL1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Auxiliary Feed Pump Room 21 ft Figure LICENSE TERMINATION PLAN Survey Package A1800 2-63

MYAPC License Termination Plan Revision 3 October 15, 2002 N

7 11 8

3 2

9 4 13 1

9 3

5 3 10 12 10 UP 2

4 4

2 11 6 14 8 HV-9 5 7 1

12 6 5 25 Purge Air Supply Unit 6 13 1

2 HV-7 1 6 8 1 3

14 Spray Pump Area 4 Heating & Ventilation 15 Unit 5 7 5

6 16 24 2 17 23 4 18 22 3 19 21 20 XX Walls XX Floor XX Equipment XX Ceiling MAINE YANKEE ATOMIC POWER CO. Site Characterization Equipment Access Area 21 ft Figure LICENSE TERMINATION PLAN Survey Package A1900 2-64

MYAPC License Termination Plan Revision 3 October 15, 2002 N

5 4 6

3 7

2 8

1 8 7 9

CPU-23 HSI-M-50 1

SIA-A-56 HSI-52 LSI-M-41 6 LSI-M-40 SIA-A-57 10 8 HSI-M-51 2

HSI-54 HSI-53 CS-65 3

CS-72 CS-66 CS-68 1 1 3 2 6 3

3 CS-73 1 4 1 5 CS-M-71 5 7

CS-67 2

4 2 2 1 3 2 4 5 XX Greenhouse Interior Walls 02WS1 3 4 5 6 XX Greenhouse Floor 02FL1 XX Equipment 02EQ1 XX Greenhouse Ceiling (plastic) 02CL1 XX Greenhouse Ceiling (I-Beams) 02CL2 XX Tank Base 01WE1 XX Greenhouse Exterior Walls 02WE1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Reactor Water Storage Tank Figure LICENSE TERMINATION PLAN Survey Package A2100 2-65

MYAPC License Termination Plan Revision 3 October 15, 2002 N

8 9 10 5 6 7 4

7 4 8 8 5

11 7

9 4 3 5 3 3 2

6 9

6 3 6 "A" BWST "B" BWST 10 6

2 2

2 12 4 7

6 1 10 5

5 11 8 1 5 1 13 1 10 SUMP 4 9 SUMP 2

4 1

11 3 11 3 2 1 S 13 12 S

36'-0" FHB XX Wall XX Floor XX Equipment XX Sump MAINE YANKEE ATOMIC POWER CO. Site Characterization Borated Water Storage Tanks Figure LICENSE TERMINATION PLAN Survey Package A2200 2-66

MYAPC License Termination Plan Revision 3 October 15, 2002 N

5 5

4 6

4 3

6 3 7 21'-0" FHB 7

8 2 2 4

3 1 1 8

3 4 2

2 2 1

1 1

XX Tank Base 01SS1 XX Shed Wall Inside 02WS1 XX Equipment 02EQ1 XX Tank Outside 01WE1 XX Shed Ceiling 02CL1 Note: 2 points also taken on shed roof, not shown 02RF1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Primary Storage Tank Area Figure LICENSE TERMINATION PLAN Survey Package A2300 2-67

MYAPC License Termination Plan Revision 3 October 15, 2002 N

XX 02SS1 XX 02WE1 XX 02EQ1 5

4 1 2 2 1 5

8 3 3 3

4 2 TT-B 4

7 6 5 5 4

1 1

2 3

1 5 7 8

3 6

1 TT-A 5

4 3 2 2

4 1 XX 01SS1 XX 01WE1 XX 01EQ1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Test Tanks 21 ft Figure LICENSE TERMINATION PLAN Survey Package A2400 2-68

MYAPC License Termination Plan Revision 3 October 15, 2002 N

3 2 BK 2 5 7 6 1 2 1

BK BK 1

2FW BK 4 5 BK 3

3 8

BK 1 3 BK 5 3

3FW 5

4 2 BK 1 4 BK 6

SOS DESK 02FL1 5 BK ALCOVE 4

9 4

1FW 2

4 PSS OFFICE NSE OFFICE 3 7

1 2 SECURITY 01FL1 XX Walls XX Floor XX Ceiling XX Equipment BK Background XFW Cable Tunnel 03FW1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Control Room & Computer Room Figure LICENSE TERMINATION PLAN Survey Package B0200 2-69

MYAPC License Termination Plan Revision 3 October 15, 2002 N 2 1

3 4 1

3 2 7 3

3 2

2 4

1 1 5 6

2 2

1 6

7 5

4 8 5 XX Walls 01WS1 XX Floor 01FL1 XX Ceiling 01CL1 XX Equipment 01EQ1 Sediment Sample X Floor Drain 01FD1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Fire Pump House Figure LICENSE TERMINATION PLAN Survey Package B0400 2-70

MYAPC License Termination Plan Revision 3 October 15, 2002 6 7 N 7 6 1 9

8 8

5 10 15 7 11 5

9 3 6 4

4 8

2 5

3 2

12 13 14 1 9

10 2

1 2 4

1 10 3

3 4 6 5

XX Cable Trays 01CL1 XX Ceiling 01EQ1 XX Stairways 01WS1, 01WS2, ... 01WS6 XX Horizontal Supports 01CL2 MAINE YANKEE ATOMIC POWER CO. Site Characterization Turbine Building 21 ft Figure LICENSE TERMINATION PLAN Survey Package B0500 2-71

MYAPC License Termination Plan Revision 3 October 15, 2002 19 20 21 22 23 11 N 7 24 10 25 BKG 5 9

18 8

26 12 13 17 27 BKG 6 28 16 15 14 29 14 01WS2 8 Sample Points 15 30 6

BKG 1 31 7

13 5 32 33 12 BKG 3 1 3 2 BKG 2 11 4

4 3 2 1 5

10 6

8 7 9

01WS3 5 Sample Points XX Walls XX Floor MAINE YANKEE ATOMIC POWER CO. Site Characterization Turbine Building 21 ft Figure LICENSE TERMINATION PLAN Survey Package B0500 2-72

MYAPC License Termination Plan Revision 3 October 15, 2002 N

9 2 1 3 Survey Unit 05 3

5 Maintenance 6 8 7 Survey 8 SW1 Unit 07 7 1 3 1

Survey RP 6 5 2 2 10 4 9

Checkpoint Unit 06 RCA 2 1 05EQ2 05EQ1 Boundary 1 4 3 Plant 2 1 2 3 3 Services 1 1 11 1 12 1

4 13 5

1 Women's 1 6 Lock Room 8 1 10 4 1 05EQ3 2 4 3 2 2 1 3 Lunch Room 1

1 3

Dosimetry 7 03SW1 Survey 4

5 RP Offices Unit 03 6 12 1 10 6 EQ2 5 1S 04EQ2 1 11 1D EQ1 10 11 7 4 04EQ1 8

1 9 6 3

4 Stock Room 9

6 Survey 8

Unit 04 5

5 2

7 Men's 4 Locker Room Survey Unit 02 6 3 9 4 3 5 10 5 4 1

2 4 4 8 1 1 3 Upstairs 2

7 Survey 2 1 2 Unit 08 3

2 3

1 1

3 2 6 XX Walls XX Floor XX Ceiling XX Equipment XX Sink XX Drain MAINE YANKEE ATOMIC POWER CO. Site Characterization Service building Cold Side 21 ft Figure LICENSE TERMINATION PLAN Survey Package B0700 2-73

MYAPC License Termination Plan Revision 3 October 15, 2002 5 6 1

4 5 4

3 7 2

2 3

EQ-1 5

1 1 2

EQ-2 1 EQ-2 4 6 EQ-4 15 EQ-4 14 8 1

10 7 EQ-3 1

1 bk 9 3

13 2 XX Walls 9 10 XX Floor 12 8 2 11 XX Ceiling XX Equipment Survey Unit 01 bk Background MAINE YANKEE ATOMIC POWER CO. Site Characterization Fuel Oil Storage Building Figure LICENSE TERMINATION PLAN Survey Package B0800 2-74

MYAPC License Termination Plan Revision 3 October 15, 2002 N

"B" Survey Unit 02 "A" Survey Unit 01 7 8 7

3 4 6 2 bk bk bk 6 bk 3

5 bk 5 9 2 1 5 8 4 bk 1 1

5 bk 3 4 bk 4

5 4

bk 6 4 bk 3 6 3 bk bk 2 2 5 bk 2 4 3 1 bk 3

9 bk 1

2 bk bk 2 5 4 1 2 3

1 bk bk 1 bk bk XX Walls XX Floor XX Ceiling XX Equipment bk Background MAINE YANKEE ATOMIC POWER CO. Site Characterization Diesel Generators Figure LICENSE TERMINATION PLAN Survey Package B0900 2-75

MYAPC License Termination Plan Revision 3 October 15, 2002 N

bk bk 3

8 EQ-5 1 7 bk bk 1 2 1

5 6 EQ-4 4

bk EQ-3 4 5 1 bk 1

bk bk 1

3 EQ-2 6 2 EQ-1 3 2 1 4

bk 1 2 5

bk bk XX Walls XX Floor XX Ceiling XX Equipment bk Background MAINE YANKEE ATOMIC POWER CO. Site Characterization Auxiliary Boiler Room Figure LICENSE TERMINATION PLAN Survey Package B1000 2-76

MYAPC License Termination Plan Revision 3 October 15, 2002 N

1 7 8 1

2 9 6 7

10 6 3

2 4

11 8 4 5

4 12 5

10 9

13 14 3

5 3 2

15 1 XX Walls XX Floor XX Equipment MAINE YANKEE ATOMIC POWER CO. Site Characterization Recirc Water Pump House Figure LICENSE TERMINATION PLAN Lower Elevation Survey Package B1100 2-77

MYAPC License Termination Plan Revision 3 October 15, 2002 8 8 7 6 5 N 4 4 9 3 6

5 3

5 3

10 7 3 2 2 2

2 10 1

5 1

11 1 1

12 13 9 14 15 XX Walls 02WS1 XX Floor 02FL1 XX Equipment 02EQ1 XX Ceiling 02CL1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Recirc Water Pump House Figure LICENSE TERMINATION PLAN Upper Elevation Survey Package B1100 2-78

MYAPC License Termination Plan Revision 3 October 15, 2002 N

6 3 bk bk 5 1

bk 1 bk bk 2

11 2 10 bk bk bk 9 bk 8 bk 12 6

1 bk 4 15 bk 3 7

3 2

8 5

10 7 4 4

2 1

9 13 14 14 11 13 3

12 15 XX Walls 01WS1 XX Floor 01FL1 XX Equipment 01EQ1, Q2, Q3 XX Ceiling 01CL1 bk Background MAINE YANKEE ATOMIC POWER CO. Site Characterization Administration Building Figure LICENSE TERMINATION PLAN Front office Survey Package B1100 2-79

MYAPC License Termination Plan Revision 3 October 15, 2002 N

12 9 4 5 3 5

2 6

3 1 2 4 4 1

2 1 7 8

10 11 3

8 7 6 First Floor / I & C Shop XX Walls 01WS1, 02WS1, 03WS1 XX Floor 01FL1, 02FL1, 03FL1 XX Equipment XX Ceiling 01CL1, 02CL1, 03CL1 MAINE YANKEE ATOMIC POWER CO. Site Characterization New Office Building (WART Bldg.) Figure LICENSE TERMINATION PLAN Survey Package B1300 2-80

MYAPC License Termination Plan Revision 3 October 15, 2002 N 12 96 11 10 9 8 7 6 5 4 3 2 1 85 N 84 73 M 15 72 15 61 L 60 49 K 48 14 37 J 36 25 H 24 13 G 12 12 1 F C 9 7 30 25 E 11 B 6 12 4 24 19 D 14 A 3 2 1 18 13 C 13 15 12 7 B 12 13 9 11 1 A 11 6 5 4 3 2 1 10 9

10 7

10 8

7 7

8 8

4 29 6

14 6

4 6

5 4

9 5 5 30 3

2 3 3 1

2 2

26 28 16 27 1

23 1

22 24 25 13 21 18 20 19 XX Walls 01WS1, 02WS1, 03WS1 XX Floor 01FL1, 02FL1, 03FL1 17 XX Equipment XX Ceiling 01CL1, 02CL1, 03CL1 Area Where Carpet Removed MAINE YANKEE ATOMIC POWER CO. Site Characterization Visitor and Information Center Figure LICENSE TERMINATION PLAN Survey Package B1400 2-81

MYAPC License Termination Plan Revision 3 October 15, 2002 7 19 6 5 4 25 N 8 18 15 14 11 9 8

3 26 2

9 7 24 6

13 12 10 17 16 20 10 5 1

11 4

21 28 25 1 12 22 13 23 24 21 2

22 3

30 16 29 26 23 27 14 30 29 17 27 20 28 15 18 19 XX Floor 01FL1 XX Ceiling 02WC1 XX Walls 02WC1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Warehouse 2 Figure LICENSE TERMINATION PLAN Survey Package B1500 2-82

MYAPC License Termination Plan Revision 3 October 15, 2002 4 3 N

6 11 3 10 2 3

4 2

12 5

7 8 1 9 9

10 14 1 13 2 1

14 11 13 7 2 1

4 8

5 6 3 15 6

5 4

12 5 XX Floor 01FL1 XX Equipment 01EQ1 XX Walls 01WS1 XX Ceiling 01CL1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Training Annex Figure LICENSE TERMINATION PLAN Survey Package B1600 2-83

MYAPC License Termination Plan Revision 3 October 15, 2002 N

7 17 24 8 1

9 107 8 21 UP 18 6 24 23 10 30 25 5 29 16 14 106 124 111 105 110 26 11 4 3 14 13 23 15 1

25 28 12 119 19 12 13 17 18 27 8 6 114 22 9 10 11 26 121 28 27 123 4 1 16 19 2 21 2 UP 30 5 20 7 29 103 116 1

15 120 118 DN XX Floor 01FL1 XX Ceiling 01CL1 XX Walls 01WS1 XX Stairwell 01SW1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Staff Building 1st Floor (22ft) Figure LICENSE TERMINATION PLAN Survey Package B1700 2-84

MYAPC License Termination Plan Revision 3 October 15, 2002 5

N 9

15 3

4 1 8 5

10 5

3 2 Platform Over 14 Tank 11 4

2 4

1 7

13 3 12 1

2 XX Equipment 6

XX Walls XX Ceiling MAINE YANKEE ATOMIC POWER CO. Site Characterization Spare Generator Building Figure LICENSE TERMINATION PLAN Survey Package B1800 2-85

MYAPC License Termination Plan Revision 3 October 15, 2002 N

4 6

5 5

4 6 1 1

2 3

3 2

XX Floor 02FL1 2 2

XX Fume Hood (All internal surfaces) 1 02MO1 XX Walls 02WC1 1

XX Ceiling 03WC1 1

Attic Floor 03Fl1; Stairway 02SW1 Bathroom Floor and Walls 02FW1 Ground Floor Walls and Ceiling 02WC1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Environmental Services Building Figure LICENSE TERMINATION PLAN Survey Package B1900 2-86

MYAPC License Termination Plan Revision 3 October 15, 2002 N

2 3

Oil Tank 3

2 2 1

Heating Unit 1

1 XX Floor 01FL1 XX Soil Sample XX Walls 01WC1 MAINE YANKEE ATOMIC POWER CO. Site Characterization Environmental Services Building Figure LICENSE TERMINATION PLAN Basement Survey Package B1900 2-87

MYAPC License Termination Plan Revision 3 October 15, 2002 N

01FL1 10 12 9

3 11 8

10 9

4 14 11 3

13 6 6 2 1 8 5

7 4

5 7 4 3

3 2

1 1

2 2

XX Floor 1 XX Background XX Walls XX Ceiling MAINE YANKEE ATOMIC POWER CO. Site Characterization Bailey Barn Figure LICENSE TERMINATION PLAN Survey Package B2000 2-88

MYAPC License Termination Plan Revision 3 October 15, 2002 N

bk 2 4 bk bk 1 bk 1

bk bk 1 bk 3

EQ-2 2

4 2

bk EQ-1 bk 1

bk bk 3 1 bk XX Ceiling XX Floor XX Equipment XX Walls bk Background MAINE YANKEE ATOMIC POWER CO. Site Characterization Lube Oil Storage Room 21 ft Figure LICENSE TERMINATION PLAN Survey Package B2100 2-89

MYAPC License Termination Plan Revision 3 October 15, 2002 N

2 bk bk 6 1 bk 1

bk EQ-1 bk 2 1 3 5 bk 4

3 bk 4

EQ-2 EQ-3 3 1 bk 1 bk 2

4 1 1 bk bk EQ-4 6 bk 5 XX Ceiling XX Floor XX Equipment XX Walls bk Background MAINE YANKEE ATOMIC POWER CO. Site Characterization Cold Machine Shop Figure LICENSE TERMINATION PLAN Turbine Building 21 ft Survey Package B2200 2-90

MYAPC License Termination Plan Revision 3 October 15, 2002 N

6 5 7 9 4 8 1

01SW2 4 5 3

2 3 1

1 2 1

6 11 01SW1 7 10 XX Floor 01FL1 12 8 XX Walls 01WS1 XX Stairwell MAINE YANKEE ATOMIC POWER CO. Site Characterization Staff building Tunnel Figure LICENSE TERMINATION PLAN Survey Package B2400 2-91

MYAPC License Termination Plan Revision 3 October 15, 2002 N

5 1 01EQ3 4

01EQ4 1 2

3 01EQ5 1

01EQ1 1 6 1

XX Ceiling XX Equipment MAINE YANKEE ATOMIC POWER CO. Site Characterization Staff building Tunnel Figure LICENSE TERMINATION PLAN Survey Package B2400 2-92

MYAPC License Termination Plan Revision 3 October 15, 2002 04WS1 FIRST AID Northwest Stairwell Exit Switchgear Room 01WS1 01FL2 01FL1 Mechanical Equipment Garage Exit Background Survey Floor Mechanical Equipment Room East Tower Exit MAINE YANKEE ATOMIC POWER CO. Site Characterization CMP Building Augusta Figure LICENSE TERMINATION PLAN Survey Package B9800 2-93

MYAPC License Termination Plan Revision 3 October 15, 2002 14 13 12 11 10 9 8 7 6

15 5 First Aid 01WS1 4 16 3 2

1

Background

Wall Survey Switchgear Room Corridor XX Approximate location of survey point.

MAINE YANKEE ATOMIC POWER CO. Site Characterization CMP building Augusta Figure LICENSE TERMINATION PLAN Survey Package B9800 2-94

MYAPC License Termination Plan Revision 3 October 15, 2002 16 15 14 13 17 12 18 19 20 21 Switchgear Background Survey Tile Walls First Aid/Bathroom 04WS1 First Aid Switchgear Room XX Approximate location of survey point.

MAINE YANKEE ATOMIC POWER CO. Site Characterization CMP Building Augusta Figure LICENSE TERMINATION PLAN Survey Package B9800 2-95

MYAPC License Termination Plan Revision 3 October 15, 2002 34 35 33 10 9 1 10 17 27 28 32 8 18 28 29 30 11 2 31 9 26 01FL2 29 30 Mechanical 7 12 19 Equipment 25 31 26 27 3 32 5 6 8 24 4 20 17 23 24 16 23 25 4 7 13 20 21 22 3 21 18 19 5 6 22 16 15 2

14 14 13 12 11 1 15 Floor- Garage Exit Background Survey XX Approximate location of survey point.

MAINE YANKEE ATOMIC POWER CO. Site Characterization CMP Building Augusta Figure LICENSE TERMINATION PLAN Survey Package B9800 2-96

MYAPC License Termination Plan Revision 3 October 15, 2002 21 21 20 18 19 22 23 27 24 25 26 32 31 30 29 28 Mechanical Equipment Room Background Survey XX Approximate location of survey point.

MAINE YANKEE ATOMIC POWER CO. Site Characterization CMP Building Augusta Figure LICENSE TERMINATION PLAN Survey Package B9800 2-97

MYAPC License Termination Plan Revision 3 October 15, 2002 26 25 24 23 22 2nd Floor Men's Bathroom Tile Wall Background Survey 04WS1 XX Approximate location of survey point.

MAINE YANKEE ATOMIC POWER CO. Site Characterization CMP Building Augusta Figure LICENSE TERMINATION PLAN Survey Package B9800 2-98

MYAPC License Termination Plan Revision 3 October 15, 2002 26 25 24 23 22 2nd Floor Men's Bathroom Tile Wall Background Survey 04WS1 XX Approximate location of survey point.

MAINE YANKEE ATOMIC POWER CO. Site Characterization CMP Building Augusta Figure LICENSE TERMINATION PLAN Survey Package B9800 2-99

MYAPC License Termination Plan Revision 3 October 15, 2002 BK 1 11 BK 10 2 9 3 4 5 6 7 8 3rd Floor Women's Bathroom Tile Wall Background Survey XX Approximate location of survey point.

MAINE YANKEE ATOMIC POWER CO. Site Characterization CMP Building Augusta Figure LICENSE TERMINATION PLAN Survey Package B9800 2-100

MYAPC License Termination Plan Revision 3 October 15, 2002 32 31 30 29 28 27 4th Floor Men's Bathroom Tile Wall Background Survey 04WS1 XX Approximate location of survey point.

MAINE YANKEE ATOMIC POWER CO. Site Characterization CMP Building Augusta Figure LICENSE TERMINATION PLAN Survey Package B9800 2-101

MYAPC License Termination Plan Attachment 2F Revision 3 Page 1 of 19 October 15, 2002 ATTACHMENT 2F Analysis of Concrete Sample Variance

MYAPC License Termination Plan Attachment 2F Revision 3 Page 2 of 19 October 15, 2002 Concrete Core Data Variance Analysis Introduction A series of concrete core samples were collected and analyzed1 as described in Engineering Calculation 011-01(MY) to determine the radionuclide mixture to use in the DCGL calculation for contaminated concrete and other contaminated materials. The nuclide mixture determination included an analysis of the data to ensure that the established dose criterion will be satisfied with sufficient confidence when the selected mixture was used. This analysis was performed primarily on the basis of dose.

This attachment describes the process used to evaluate the nuclide mixture for contaminated concrete surfaces and to determine that the mixture is representative and ensures that the established dose criterion will be met with sufficient confidence.

Nuclide Data The concrete core data used to determine the nuclide mixture was collected during two sampling campaigns. The first data set was comprised of seven cores collected during the site characterization that were representative of concrete contamination in the majority of plant areas. This majority area is called the balance of plant (BOP). The first data set was used to determine the nuclide fractions for the BOP, which includes most of the areas in the building basements.

A second data set, consisting of eight samples, was collected to replace cores consumed during analysis processes, to investigate suspect data, and to provide additional information on the nuclide mixture in certain areas that had some potential for containing nuclide mixtures that differed from the BOP. The second data set consisted of two cores from the Containment Outer Annulus (O/A) trench, three cores from within the loops of Containment, and three cores from the Primary Auxiliary Building (PAB). See Tables 1 and 5 for the listing of actual core identification numbers and the associated plant locations. (Location maps are included in Engineering Calculation EC-011-01 (MY).)

Conversion of BOP Concrete Core Analytical Results to Dose The first step in determining the acceptability of the BOP nuclide mixture was to normalize2 the nuclide data and convert the normalized data to dose. Dose was used in the evaluation since the unrestricted use criterion is defined in terms of dose and expressing potential uncertainty in terms of dose provides the most direct means of demonstrating acceptability.

There were several steps required to convert the raw radioanalytical core data to dose. First, the nuclide data for each core was decay corrected to 1/1/2004 to correspond to the approximate 1

Core analyses were performed by Duke Engineering and Services Environmental Laboratory 2

Normalization, in this case refers to converting the reported nuclide concentration results into nuclide fractions.

MYAPC License Termination Plan Attachment 2F Revision 3 Page 3 of 19 October 15, 2002 time of the last final surveys. The initial and decay corrected data as well as other supporting documents is provided in EC-011-01. Second, the decay corrected nuclide concentration results from each of the cores were converted to fractions. The sum of the nuclide fractions in each core then represent 1.0 dpm/100 cm2 total activity. Analytical results that were reported as less than the minimum detectable activity (MDA) were assumed to be present at the MDA value in the initial review. The nuclides that were listed as less than MDA in each of the seven cores are indicated in Table 1 by a < sign. See Table 1, Column 3 for example of normalized nuclide fractions for the core 1FL1.

The basement fill model (LTP Section 6.6.1) was used to convert the normalized nuclide fractions to dose. Note that there were two other materials, i.e., buried pipe/conduit and embedded pipe, that were assumed to contain the BOP nuclide mixture and each of these materials has a different dose model. However, because the basement concrete contains the overwhelming majority of the contamination inventory and results in the highest dose, the basement fill model was selected for the core dose calculations.

The dose was calculated by multiplying the normalized nuclide fractions by the unitized dose factors determined in Engineering Calculation EC-011-01 (MY). The unitized dose factor is the dose that would result from 1.0 dpm/100 cm2 activity of a given radionuclide. See Table 1, Column 4 for an example of the dose from the nuclide fractions in the Core 1FL1 mixture.

The sum of the normalized doses from all radionuclides in a given core represents the dose from each core assuming that the core contains a total activity of 1.0 dpm/100 cm2. The last conversion required to perform the analysis of uncertainty in the radionuclide mixture is to convert the 1.0 dpm/100 cm2 normalized doses to a dose that represents 18,000 dpm/100 cm2 detectable beta activity. This is accomplished by dividing each of the nuclides in a given core by the detectable beta fraction of the core and multiplying by 18,000 dpm/100 cm2. This conversion allows direct comparison with the dose that would result if residual contamination were present in each core at the DCGL concentrations of 18,000 dpm/100 cm2 observable data.

See Table 1, Column 5 for an example of the nuclide dose from core 1FL1 after converting to 18,000 dpm/100 cm2.

The use of the various dose-converted core data sets in the evaluation of core variability is described in the sections below.

Evaluation of Less than MDA Nuclides Before the nuclide data variability could be evaluated, the results reported as less than MDA were considered. It was expected that several of the 31 nuclides would be reported as less than MDA since these nuclides have a low probability of being present and were included in the analyses only as a conservative measure. Two approaches were considered for evaluating MDA results; 1) include the MDA values as representing actual concentrations, and 2) remove the non-detected nuclides from the mixture. Removing the nuclides was considered more appropriate and representative of actual site conditions because the non-detected nuclides are believed either

MYAPC License Termination Plan Attachment 2F Revision 3 Page 4 of 19 October 15, 2002 to not be present or to be present at concentrations well below the reported MDA value.

However, it cannot be ruled out with 100% certainty that the non-detected nuclides are not present at activities approaching the MDA values. Therefore, an analysis was performed, based on relative dose, to review the affect of leaving the MDAs in the mixture versus removing the MDAs.

To perform this analysis, the dose from the MDA nuclides was compared to the total dose including all nuclides. A nuclide was included in the MDA category if it was not detected in any of the cores. If a nuclide was detected in one or more cores, the nuclide was retained and included in the mixture calculation, including MDA values in some instances. For example, Sr-90 was detected above MDA in three of the seven primary cores. For the remaining 4 cores, the MDA value was conservatively assumed to represent detectable activity. As shown in Table 2, the MDA nuclides contributed 1.8% of the total dose (5.1E-03 mrem/y/2.8E-01mrem/y). Since the MDA contribution was low the MDA nuclides were removed from the mixture.

Table 3 contains the nuclide mixture after the MDA nuclides were removed. Note that the nuclide fractions listed in Table 3 was renormalized to 1.0 after removal of the MDA radionuclides that were not detected in any of the cores. This is a conservative yet appropriate approach since the MDA radionuclides were not believed to be present in appreciable quantities.

Evaluation of Variability of Dose from Primary Seven Core Data Set The variability of the dose from the cores in the primary seven core data set, after removal of MDAs, was evaluated to demonstrate that the variability is low relative to the unrestricted use dose criteria of 10 mrem/yr all pathways and that the seven core data set is sufficiently representative of BOP areas. The variability was evaluated by reviewing the dose from individual cores and the dose from the average of the nuclide fractions. The mean and standard deviation of the dose from both the individual cores and the nuclide fractions were evaluated to determine: 1) if there were a significant difference in the means calculated using the two methods, 2) whether any individual core dose appeared to be significantly different from the mean dose, and 3) whether the variability of the mean dose using the average of the fractions method was sufficiently low relative to the 10 mrem/yr all pathways unrestricted use criteria to provide confidence that the dose criterion would be satisfied using the average of the fractions method.

Calculation of Mean and Standard Deviation The mean dose and standard deviation of the mean from the individual seven cores were calculated using the data set generated after removal of MDAs and converting to 18,000 dpm measurable gross beta (See Table 4). The calculation of the standard deviation of the mean from the individual cores used the following standard equations:

n x 2 ( x ) 2 n(n 1)

MYAPC License Termination Plan Attachment 2F Revision 3 Page 5 of 19 October 15, 2002 then, Standard Deviation n

The mean dose and standard deviation of the mean from the nuclide fractions in the seven cores were calculated using the data set generated after removal of MDAs but before converting to 18,000 dpm/100 cm2 measurable gross beta. Use of this data set was required because the relative nuclide fractions found in the original core analyses need to be retained to correctly calculate the average of the nuclide fractions over the seven cores. After the dose from average nuclide fractions was calculated the result was converted to represent the dose from 18,000 dpm/100 cm2 measurable gross beta prior to comparison of the two data sets.

The mean dose from the nuclide fractions was calculated by summing the dose from average of each nuclide fraction over the seven cores. The standard deviation of the mean dose from the nuclide fractions required the use of a standard propagation of errors equation to account for the variability within each average nuclide fraction. This was accomplished by squaring the standard deviation of each average nuclide fraction and summing over all nuclides. The propagated error was calculated as:

2 std. dev.

n The first and second data sets were evaluated and compared to ensure that there was not a significant variation between the average dose from the individual cores, which is assumed to represent a given area of the plant, and the average dose as represented by the nuclide fractions in the BOP mixture listed in the LTP.

Data Evaluation The first evaluation of the individual core and nuclide fractions data sets was performed to demonstrate that there was no significant difference between the means of the two methods for calculating mean dose. It is obvious by a simple comparison of the means and standard deviations provided in Table 4 that there is not a significant difference between the means. The mean dose and standard deviation for the individual cores are 0.29 mrem/yr and 0.030 mrem/yr, respectively. The mean dose and standard deviation using the average of the fractions method are 0.30 mrem/yr and 0.070 mrem/yr, respectively. In fact, the means are essentially identical, differing by less than 0.01 mrem/yr.

The second evaluation entailed a review of the individual core data set to determine if any individual core dose was significantly different from the mean. The standard deviation of the individual core dose, 0.083 mrem/yr, was used for this evaluation where: Table 4, individual core standard deviation of the mean (0.031) is multiplied by the square root of the sample

MYAPC License Termination Plan Attachment 2F Revision 3 Page 6 of 19 October 15, 2002 population (2.65). Multiplying the standard deviation by 1.96 and then adding and subtracting the result to the mean results in the upper and lower 95% confidence level bounds. The upper confidence level was 0.46 mrem/yr and the lower confidence level was 0.13 mrem/yr. No individual core dose was outside the 95% confidence levels indicating that no area represented by the core dose was significantly different from the mean. Note that core 02FL51 was at the upper confidence level. This is attributable to an unusually high MDA value for Sr-90 in this core relative to the Sr-90 MDA values reported for the 4 other cores where MDA values were applied and is therefore not significant.

The third evaluation was to review the distribution of the mean from the average of the fractions method. As seen in Table 4, upper 95% confidence level is 0.14 mrem/yr (0.07 mrem/yr times 1.96), which is a very small fraction of the of the 10 mrem/yr dose criterion.

The results of the three evaluations performed above demonstrate: 1) that there is no significant difference between the individual core and average of the fractions methods for calculating mean dose, 2) that no individual core varied significantly from the mean indicating that all of the cores were a part of the same population, and 3) that the variability of the dose using the average of the fractions method is a small fraction of the unrestricted use limit and ensures that the dose criterion will be met with sufficient confidence.

Methods for Evaluating Additional Eight Cores The discussions and analyses presented above demonstrate that the seven core data set is sufficient to determine the BOP nuclide mixture. The next task was to develop the methods to evaluate the nuclide mixtures in the eight additional cores that were collected during continuing characterization and determine whether they were consistent with the BOP mixture.

If the nuclide mixture of a given core is significantly different from the BOP mixture, then a separate mixture and DCGL may be necessary for the areas represented by the cores. These evaluation criteria would also apply to additional concrete cores collected, if any. Based on evaluation of the 15 cores and a review of the potential for additional plant areas to have a significantly different nuclide mixture than the BOP, no additional cores are deemed necessary to support the LTP.

Three factors were considered in the evaluation of additional cores: 1) whether the core contained detectable transuranics, 2) whether one or more radionuclide fractions are significantly different from the BOP mixture, and 3) whether the dose from an additional core was significantly different from the BOP mixture dose and exceeded 1.0 mrem/yr. The three evaluation factors were developed during the Technical Issue Resolution Process (TIRP) conducted by the State of Maine and Maine Yankee as a part of the Settlement Agreement related to the States motion to terminate their petition to intervene in the matter of MYs proposed LTP. During the TIRP, MY and State technical experts developed and used these three criteria to evaluate additional concrete core samples. Maine Yankee believes the criteria are reasonable and protective and agreed to include the criteria in the LTP. The three criteria for

MYAPC License Termination Plan Attachment 2F Revision 3 Page 7 of 19 October 15, 2002 evaluating individual cores are listed below.

1. No detectable TRU.

2. Individual fractions of nuclides:

Nuclide Maximum Nuclide Fraction Sr 0.013 Co 0.170 Cs 1.000 Ni 1.000

3. Individual core total dose from all nuclide fractions less than 1.0 mrem/yr.

The first individual core criterion (#1 above) pertained to transuranic (TRU) radionuclides. The TRUs were singled out because their radiological and chemical characteristics differ from the BOP radionuclide mixture, as well as the fact that there is a significant level of stakeholder interest in TRUs. Therefore, the first individual core decision statement was whether or not the core contained TRUs at levels exceeding the minimum detectable activity (MDA). If so, the area represented by the core would either 1) be subject to a unique radionuclide mixture and DCGL or 2) be combined with other TRU cores to generate a single radionuclide mixture and DCGL representing several TRU affected areas.

The second individual core criterion (#2 above) compares the radionuclide fractions in a given core to an upper bound expected given the data provided in the seven-core BOP set. The upper 95% confidence level (UCL) was calculated for each nuclide fraction in the seven core set. The UCLs for four nuclides, Cs-137, Co-60, Sr-90, and Ni-63 are listed above as individual core Criterion #2. Criterion #2 was limited to these four nuclides since they together comprise the overwhelming majority of the dose from concrete basement surfaces. If the nuclide fractions in a given individual core are all less than the values listed in Criterion#2, the BOP radionuclide fraction set is assumed to sufficiently represent the core. However, if an individual core contains a nuclide fraction equal to or exceeding one of the values listed in Criterion #2, then the dose from the core must be calculated and compared to the dose listed in individual core Criterion #3

(#3 above). Criterion # 2 only applies to nuclide fractions that are based upon nuclide activities greater than MDA. If a core's radionuclide fraction, which is based upon a radionuclide activity less than MDA, fails to meet Criterion #2 and the MDA is comparable to the MDA's achieved for the other cores, then it will be considered as having satisfied Criterion #2 for that radionuclide.

The third individual core criterion (#3 above) ensures that the dose potentially represented by an individual core is not significantly different from the seven-core data set. Criterion #3 is required only if Criterion #2 is not satisfied. The dose criterion of 1.0 mrem/yr was selected because at the time of the TIRP original consensus 1.0 mrem/yr was 0.44 mrem/yr above the mean dose (0.556 mrem/yr) calculated using the BOP nuclide mixture. Using the most current dose

MYAPC License Termination Plan Attachment 2F Revision 3 Page 8 of 19 October 15, 2002 assessment results, 1 mrem/yr is 0.70 mrem/y above the mean dose of 0.30 mrem/y. The current 0.70 mrem/yr value is more conservative than the 0.44 mrem/yr value found to be acceptable by the TIRP since the actual core variability is a smaller percentage of the acceptable 0.70 mrem/yr variability. A variability of either 0.44 or 0.70 mrem/y above the mean is well below the value that would be acceptable by NRC guidance in NUREG-1727, Page E16, which states that the presence of nuclides that likely contribute less than 10% of the total effective dose equivalent may be ignored. The Maine Yankee dose limit is 10 mrem/yr and 10% equals 1.0 mrem/yr. Use of such a 10% criterion is also supported by NRC regulations in 10 CFR 20.1204(g) and 10 CFR 20.1502. Note that the value of 0.70 mrem/yr represents the variability attributable to uncertainty in the nuclide mixture for the individual concrete cores and not an actual dose above the 10 mrem/yr limit. The best estimate of dose is calculated using the mean nuclide fractions of the seven-core data set, i.e., the BOP nuclide mixture.

Results of Evaluation of Eight Additional Cores The analytical results for the eight additional cores are provided in Table 5. The data was reduced in the same manner as the primary seven core data set. Each radionuclide was decay corrected to 2004; the nuclide fractions were normalized to 1.0; the nuclide fractions were then multiplied by the unitized dose factors based on 1.0 dpm/100 cm2 (to convert the fractions to dose); and finally, the dose was converted to that which would result from 18,000 dpm/100 cm2 measurable gross beta. The data for the individual cores was then compared to the three evaluation criteria described above.

Inspection of Table 5 shows that four of the cores clearly meet the 3 evaluation criteria and are considered to be sufficiently represented by the BOP nuclide mixture. These cores were collected from the Containment loops 1, 2, and 3, and the PAB evaporator cubicle and show good agreement with the BOP mixture.

Of the remaining four cores, two of the cores were from the O/A trench and two are from the PAB pipe tunnel. Three of the four cores contained TRUs that were above the MDA. Table 6 contains the radionuclide mixture and dose data for the four TRU affected cores after removal of the MDA results. Table 7 contains the dose summaries for the individual core and average of the fractions methods for the TRU affected cores. The individual core doses range from 0.18 to 0.25 mrem/yr assuming 18,000 dpm/100 cm2 observable beta. This is a very low fraction of the unrestricted use criteria and is much less than the 1.0 mrem/yr individual core dose criteria used in the BOP nuclide mixture decision rule. As stated previously, the data reduction method for these four cores was conducted in the same manner as for the BOP cores. The dose from the four cores TRU mixture using the average of the fractions method was 0.21 mrem/yr. Based on these results, the four core average of the fractions nuclide mixture will be used to determine a separate DCGL for TRU Affected areas hereafter referred to as Special Areas.

After the identification of Special Areas through the analyses of the eight additional cores, a review of building basement areas was performed to determine if there were other areas that could be designated as Special Areas that were not represented by the BOP mixture. The liquid

MYAPC License Termination Plan Attachment 2F Revision 3 Page 9 of 19 October 15, 2002 waste stream significantly impacted both the O/A trench and PAB pipe tunnel. The O/A trench captured all water released to the floor of the Containment building and routed it to the Containment sump. The PAB pipe tunnel held the pipes that carried the liquid waste water being processed by the filters and demineralizers in the PAB. Both areas had standing water and boron encrustations during plant operation. As a result of this review one additional area (letdown heat exchanger cubicle) was identified that had operating history and characteristics that were sufficiently similar to the PAB pipe tunnel and Containment O/Annulus Trench to warrant consideration as TRU affected.

The letdown heat exchanger cubicle is an area of approximately 2.5 m by 2.5 m by 3 m tall located in the PAB basement. Because of its small size, it was not specifically sampled. This is the one area that stands out as perhaps needing to be examined since it was processed high temperature liquids and had standing boron. Additional cores samples could have been collected in the letdown heat exchanger cubicle to demonstrate that the cubicle is not TRU affected and that the BOP nuclide mixture would apply. However, the decision was made to conservatively assume the area was TRU affected and to use the Special Area nuclide mixture to calculate the DCGL for this area. This decision is conservative since the DCGL for Special Areas is lower than the BOP areas.

Conclusion The nuclide mixture provided in Table 4 Column 2 using the average of the fractions methods has been demonstrated to be representative of BOP areas and ensures that the established dose criterion will be satisfied with sufficient confidence. Three TRU affected areas have been identified that are represented by a unique nuclide mixture as listed in Table 7 Column 2.

Finally a decision rule has been developed through the cooperative efforts of the State of Maine and Maine Yankee (i. e., TIRP). This rule will be used to evaluate the impact and use of any future core information obtained with regard to nuclide mixture and the associated DCGL.

References

1. Maine Yankee License Termination Plan Settlement Agreement, ASLNP No. 00-870-03-0LA, August 29, 2001.

2. Participant Consensus Agreement, State of Maine - Maine Yankee Settlement Agreement, Technical Issue Resolution Process, Dated December 13, 2001.

Table 1 Attachment 2F Nuclide Fractions and Dose (mrem/y) for Balance of Plant Core Samples Analysis of Concrete Sample Variance (Table 1 page 1 of 2)

Page 10 of 19 Column # ==> 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 PAB 11' West Dose For Fuel Bldg Dose For Dose For Dose For Pipe Trench 1.80E+04 Decon Room 1.80E+04 Spray Bldg 11' 1.80E+04 RCA Bldg 21' 1.80E+04 1FL1 dpm/100 cm2 01FL31 dpm/100 cm2 01FL41 dpm/100 cm2 01FL61 dpm/100 cm2 2004 1 dpm dose detectable 2004 1 dpm dose detectable 2004 1 dpm dose detectable 2004 1 dpm dose detectable Nuclide nf Times nf 2004 nf Times nf 2004 nf Times nf 2004 nf Times nf 2004 H-3 3.997E-04 1.34E-08 9.59E-04 6.265E-04 2.10E-08 4.84E-04 1.211E-03 4.06E-08 4.85E-03 1.025E-03 3.43E-08 7.08E-04 C-14 < 3.433E-06 6.19E-11 4.43E-06 < 1.067E-04 1.92E-09 4.44E-05 < 1.531E-04 2.76E-09 3.30E-04 < 3.590E-05 6.47E-10 1.33E-05 Mn-54 < 1.369E-04 2.88E-10 2.06E-05 < 1.326E-04 2.79E-10 6.43E-06 < 1.533E-04 3.22E-10 3.86E-05 < 1.098E-04 2.31E-10 4.76E-06 Fe-55 2.855E-03 1.67E-09 1.19E-04 4.571E-04 2.67E-10 6.16E-06 3.800E-03 2.22E-09 2.65E-04 2.896E-03 1.69E-09 3.49E-05 Co-57 2.861E-05 7.26E-12 5.20E-07 < 9.019E-05 2.29E-11 5.27E-07 < 9.928E-05 2.52E-11 3.01E-06 < 5.719E-05 1.45E-11 2.99E-07 Co-58 < 5.437E-10 4.40E-16 3.15E-11 < 4.137E-10 3.34E-16 7.71E-12 < 5.137E-10 4.15E-16 4.97E-11 < 3.427E-10 2.77E-16 5.71E-12 Ni-59 < 7.594E-03 9.18E-11 6.57E-06 < 2.152E-03 2.60E-11 6.00E-07 < 8.586E-03 1.04E-10 1.24E-05 < 1.223E-03 1.48E-11 3.04E-07 Co-60 1.539E-01 9.70E-07 6.95E-02 4.057E-03 2.56E-08 5.90E-04 3.125E-02 1.97E-07 2.36E-02 2.692E-02 1.70E-07 3.50E-03 Ni-63 7.357E-01 8.22E-08 5.89E-03 2.085E-01 2.33E-08 5.37E-04 8.318E-01 9.29E-08 1.11E-02 1.184E-01 1.32E-08 2.73E-04 Zn-65 < 1.324E-04 1.40E-09 1.00E-04 < 1.383E-04 1.46E-09 3.37E-05 < 2.214E-04 2.34E-09 2.80E-04 < 8.986E-05 9.50E-10 1.96E-05 Sr-90 1.316E-04 8.30E-09 5.95E-04 < 1.198E-03 7.56E-08 1.74E-03 5.420E-04 3.42E-08 4.09E-03 1.267E-03 8.00E-08 1.65E-03 Nb-94 < 5.978E-03 1.00E-08 7.17E-04 < 3.825E-03 6.40E-09 1.48E-04 < 9.049E-03 1.51E-08 1.81E-03 < 3.867E-03 6.47E-09 1.33E-04 Tc-99 < 1.520E-05 4.89E-09 3.50E-04 < 1.409E-04 4.53E-08 1.04E-03 < 1.547E-05 4.97E-09 5.95E-04 < 1.558E-04 5.01E-08 1.03E-03 Ru-106 < 1.742E-03 2.15E-08 1.54E-03 < 3.616E-03 4.45E-08 1.03E-03 < 3.178E-03 3.91E-08 4.68E-03 < 2.158E-03 2.66E-08 5.48E-04 Ag-110m < 1.003E-04 3.46E-10 2.48E-05 < 7.361E-05 2.54E-10 5.85E-06 < 1.228E-04 4.23E-10 5.06E-05 < 5.515E-05 1.90E-10 3.92E-06 Sb-125 < 3.596E-03 7.22E-09 5.17E-04 < 1.165E-02 2.34E-08 5.39E-04 < 7.372E-03 1.48E-08 1.77E-03 < 7.898E-03 1.59E-08 3.27E-04 I-129 < 3.032E-08 1.97E-10 1.41E-05 < 2.810E-07 1.82E-09 4.20E-05 < 3.084E-08 2.00E-10 2.39E-05 < 3.107E-07 2.02E-09 4.15E-05 Cs-134 1.720E-03 4.10E-08 2.94E-03 < 1.264E-03 3.01E-08 6.95E-04 < 2.039E-03 4.86E-08 5.81E-03 1.835E-03 4.38E-08 9.01E-04 Cs-137 8.049E-02 1.29E-06 9.25E-02 7.461E-01 1.20E-05 2.76E-01 8.188E-02 1.31E-06 1.57E-01 8.250E-01 1.32E-05 2.73E-01 Ce-144 < 3.222E-04 3.34E-10 2.39E-05 < 9.153E-04 9.49E-10 2.19E-05 < 9.014E-04 9.34E-10 1.12E-04 < 5.599E-04 5.80E-10 1.20E-05 Pm-147 < 5.086E-06 1.03E-11 7.39E-07 < 2.762E-04 5.60E-10 1.29E-05 < 2.629E-05 5.33E-11 6.38E-06 < 2.761E-05 5.60E-11 1.15E-06 Eu-154 < 2.749E-03 2.11E-09 1.51E-04 < 3.889E-03 2.99E-09 6.89E-05 < 9.220E-03 7.09E-09 8.48E-04 < 1.685E-03 1.30E-09 2.67E-05 Eu-155 < 2.347E-03 2.31E-10 1.65E-05 < 7.394E-03 7.27E-10 1.68E-05 < 7.950E-03 7.81E-10 9.34E-05 < 4.368E-03 4.29E-10 8.85E-06 Pu-238 < 8.588E-07 1.20E-10 8.56E-06 < 1.537E-05 2.14E-09 4.93E-05 < 1.768E-05 2.46E-09 2.94E-04 < 5.462E-06 7.60E-10 1.57E-05 Pu-239 < 4.261E-07 6.56E-11 4.70E-06 < 5.260E-06 8.10E-10 1.87E-05 < 7.773E-06 1.20E-09 1.43E-04 < 2.728E-06 4.20E-10 8.65E-06 Pu-240 < 4.259E-07 6.56E-11 4.70E-06 < 5.258E-06 8.10E-10 1.87E-05 < 7.771E-06 1.20E-09 1.43E-04 < 2.727E-06 4.20E-10 8.65E-06 Pu-241 < 6.122E-05 1.83E-10 1.31E-05 < 3.324E-03 9.92E-09 2.29E-04 < 3.164E-04 9.44E-10 1.13E-04 < 3.324E-04 9.92E-10 2.04E-05 Am-241 < 1.768E-06 8.24E-11 5.90E-06 < 3.481E-05 1.62E-09 3.74E-05 < 2.304E-05 1.07E-09 1.28E-04 < 8.914E-06 4.15E-10 8.55E-06 Cm-242 < 6.341E-10 4.44E-16 3.18E-11 < 1.225E-08 8.58E-15 1.98E-10 < 8.724E-09 6.11E-15 7.31E-10 < 3.500E-09 2.45E-15 5.05E-11 Cm-243 < 3.641E-07 5.57E-12 3.99E-07 < 5.775E-06 8.84E-11 2.04E-06 < 4.878E-06 7.47E-11 8.93E-06 < 1.823E-06 2.79E-11 5.75E-07 Cm-244 < 3.416E-07 4.21E-12 3.01E-07 < 5.418E-06 6.67E-11 1.54E-06 < 4.576E-06 5.64E-11 6.74E-06 < 1.710E-06 2.11E-11 4.34E-07 sum 1.000E+00 1.000E+00 1.000E+00 1.000E+00 obs. fraction 2.513E-01 7.809E-01 1.505E-01 8.736E-01

Table 1 Attachment 2F Nuclide Fractions and Dose (mrem/y) for Balance of Plant Core Samples Analysis of Concrete Sample Variance (Table 1 page 2 of 2) Page 11 of 19 Column # => 18 19 20 21 22 23 24 25 26 27 28 29 33 PAB 11' Dose For CTMT Dose For CTMT Dose For Pipe Trench 1.80E+04 -2' Loop 2 1.80E+04 -2' Loop 1 1.80E+04 2 2 2 01FL81 dpm/100 cm 02FL21 dpm/100 cm 02FL51 dpm/100 cm 2004 1 dpm dose detectable 2004 1 dpm dose detectable 2004 1 dpm dose detectable 1.0 dpm Nuclide nf Times nf 2004 nf Times nf 2004 nf Times nf 2004 dose factor H-3 3.583E-03 1.20E-07 2.65E-03 4.279E-03 1.43E-07 2.65E-03 1.214E-01 4.07E-06 1.26E-01 3.351E-05 C-14 < 9.586E-05 1.73E-09 3.82E-05 < 1.329E-04 2.40E-09 4.42E-05 < 2.419E-03 4.36E-08 1.35E-03 1.803E-05 Mn-54 < 1.198E-04 2.52E-10 5.56E-06 < 2.051E-04 4.32E-10 7.97E-06 < 1.356E-03 2.85E-09 8.85E-05 2.104E-06 Fe-55 2.372E-03 1.39E-09 3.06E-05 < 5.234E-04 3.06E-10 5.65E-06 1.646E-02 9.61E-09 2.98E-04 5.843E-07 Co-57 < 6.477E-05 1.64E-11 3.63E-07 < 1.157E-04 2.94E-11 5.42E-07 < 1.409E-03 3.57E-10 1.11E-05 2.537E-07 Co-58 < 4.361E-10 3.53E-16 7.78E-12 < 6.556E-10 5.30E-16 9.79E-12 < 1.135E-08 9.18E-15 2.85E-10 8.086E-07 Ni-59 < 1.792E-03 2.17E-11 4.78E-07 < 1.593E-04 1.93E-12 3.56E-08 < 2.648E-03 3.20E-11 9.94E-07 1.209E-08 Co-60 5.049E-02 3.18E-07 7.03E-03 5.989E-03 3.78E-08 6.97E-04 1.034E-01 6.52E-07 2.02E-02 6.305E-06 Ni-63 1.736E-01 1.94E-08 4.28E-04 1.543E-02 1.72E-09 3.18E-05 2.565E-01 2.87E-08 8.90E-04 1.117E-07 Zn-65 < 1.469E-04 1.55E-09 3.43E-05 < 1.508E-04 1.59E-09 2.94E-05 < 3.794E-03 4.01E-08 1.25E-03 1.058E-05 Sr-90 < 4.383E-04 2.77E-08 6.10E-04 < 3.843E-04 2.42E-08 4.48E-04 < 1.227E-02 7.74E-07 2.40E-02 6.310E-05 Nb-94 < 5.305E-03 8.88E-09 1.96E-04 < 4.703E-03 7.87E-09 1.45E-04 < 5.249E-02 8.79E-08 2.73E-03 1.674E-06 Tc-99 < 1.398E-04 4.50E-08 9.93E-04 < 1.763E-04 5.67E-08 1.05E-03 < 4.958E-05 1.59E-08 4.95E-04 3.216E-04 Ru-106 < 2.533E-03 3.12E-08 6.89E-04 < 3.326E-03 4.10E-08 7.56E-04 < 4.993E-03 6.15E-08 1.91E-03 1.232E-05 Ag-110m < 9.030E-05 3.11E-10 6.88E-06 < 7.082E-05 2.44E-10 4.51E-06 < 1.319E-03 4.55E-09 1.41E-04 3.449E-06 Sb-125 < 7.779E-03 1.56E-08 3.45E-04 < 1.364E-02 2.74E-08 5.06E-04 < 8.338E-02 1.67E-07 5.19E-03 2.007E-06 I-129 < 2.788E-07 1.81E-09 3.99E-05 < 3.515E-07 2.28E-09 4.21E-05 < 9.887E-08 6.42E-10 1.99E-05 6.489E-03 Cs-134 1.552E-03 3.70E-08 8.17E-04 1.534E-03 3.66E-08 6.75E-04 < 1.750E-02 4.17E-07 1.29E-02 2.384E-05 Cs-137 7.402E-01 1.19E-05 2.62E-01 9.332E-01 1.50E-05 2.77E-01 2.625E-01 4.21E-06 1.31E-01 1.605E-05 Ce-144 < 6.548E-04 6.79E-10 1.50E-05 < 1.209E-03 1.25E-09 2.31E-05 < 1.379E-02 1.43E-08 4.44E-04 1.036E-06 Pm-147 < 2.584E-05 5.24E-11 1.16E-06 < 8.857E-05 1.80E-10 3.32E-06 < 8.550E-04 1.73E-09 5.38E-05 2.028E-06 Eu-154 < 3.556E-03 2.73E-09 6.04E-05 < 4.704E-03 3.62E-09 6.68E-05 < 2.092E-02 1.61E-08 4.99E-04 7.690E-07 Eu-155 < 5.055E-03 4.97E-10 1.10E-05 < 8.838E-03 8.69E-10 1.60E-05 < 8.969E-03 8.82E-10 2.74E-05 9.828E-08 Pu-238 < 1.476E-05 2.05E-09 4.54E-05 < 1.045E-05 1.45E-09 2.69E-05 < 2.307E-04 3.21E-08 9.96E-04 1.392E-04 Pu-239 < 7.040E-06 1.08E-09 2.39E-05 < 4.601E-06 7.08E-10 1.31E-05 < 8.748E-05 1.35E-08 4.18E-04 1.540E-04 Pu-240 < 7.037E-06 1.08E-09 2.39E-05 < 4.599E-06 7.08E-10 1.31E-05 < 8.745E-05 1.35E-08 4.18E-04 1.540E-04 Pu-241 < 3.110E-04 9.28E-10 2.05E-05 < 1.066E-03 3.18E-09 5.88E-05 < 1.029E-02 3.07E-08 9.54E-04 2.985E-06 Am-241 < 3.241E-05 1.51E-09 3.33E-05 < 2.233E-05 1.04E-09 1.92E-05 < 5.790E-04 2.70E-08 8.37E-04 4.658E-05 Cm-242 < 1.318E-08 9.23E-15 2.04E-10 < 6.351E-09 4.45E-15 8.21E-11 < 2.622E-07 1.84E-13 5.70E-09 7.002E-07 Cm-243 < 6.663E-06 1.02E-10 2.25E-06 < 3.750E-06 5.74E-11 1.06E-06 < 1.304E-04 2.00E-09 6.20E-05 1.531E-05 Cm-244 < 6.250E-06 7.70E-11 1.70E-06 < 3.518E-06 4.33E-11 8.00E-07 < 1.224E-04 1.51E-09 4.68E-05 1.232E-05 sum 1.000E+00 1.000E+00 1.000E+00 obs. fraction 8.154E-01 9.747E-01 5.799E-01 avg. fraction 6.324E-01

Table 2 Attachment 2F Calculation of Dose From MDA Nuclides Analysis of Concrete Sample Variance Page 12 of 19 Column # ==> 2 3 4 5 6 7 8 9 10 11 Dose Results (in mrem/y) for Average of the Fractions (1.0 dpm)

Nuclide 1FL1 01FL31 01FL41 01FL61 01FL81 02FL21 02FL51 mean* stdev mean (stdev Mean)2 H-3 1.34E-08 2.10E-08 4.06E-08 3.43E-08 1.20E-07 1.43E-07 4.07E-06 1.81E-02 1.629E-02 2.65E-04 C-14 6.19E-11 1.92E-09 2.76E-09 6.47E-10 1.73E-09 2.40E-09 4.36E-08 2.16E-04 1.712E-04 2.93E-08 Mn-54 2.88E-10 2.79E-10 3.22E-10 2.31E-10 2.52E-10 4.32E-10 2.85E-09 1.89E-05 1.040E-05 1.08E-10 Fe-55 1.67E-09 2.67E-10 2.22E-09 1.69E-09 1.39E-09 3.06E-10 9.61E-09 6.98E-05 3.488E-05 1.22E-09 Co-57 7.26E-12 2.29E-11 2.52E-11 1.45E-11 1.64E-11 2.94E-11 3.57E-10 1.92E-06 1.377E-06 1.90E-12 Co-58 4.40E-16 3.34E-16 4.15E-16 2.77E-16 3.53E-16 5.30E-16 9.18E-15 4.69E-11 3.574E-11 1.28E-21 Ni-59 9.18E-11 2.60E-11 1.04E-10 1.48E-11 2.17E-11 1.93E-12 3.20E-11 1.19E-06 4.261E-07 1.82E-13 Co-60 9.70E-07 2.56E-08 1.97E-07 1.70E-07 3.18E-07 3.78E-08 6.52E-07 9.64E-03 3.765E-03 1.42E-05 Ni-63 8.22E-08 2.33E-08 9.29E-08 1.32E-08 1.94E-08 1.72E-09 2.87E-08 1.06E-03 3.816E-04 1.46E-07 Zn-65 1.40E-09 1.46E-09 2.34E-09 9.50E-10 1.55E-09 1.59E-09 4.01E-08 2.01E-04 1.569E-04 2.46E-08 Sr-90 8.31E-09 7.56E-08 3.42E-08 8.00E-08 2.77E-08 2.42E-08 7.74E-07 4.16E-03 2.992E-03 8.95E-06 Nb-94 1.00E-08 6.40E-09 1.51E-08 6.47E-09 8.88E-09 7.87E-09 8.79E-08 5.80E-04 3.217E-04 1.04E-07 Tc-99 4.89E-09 4.53E-08 4.97E-09 5.01E-08 4.50E-08 5.67E-08 1.59E-08 9.06E-04 2.408E-04 5.80E-08 Ru-106 2.15E-08 4.45E-08 3.91E-08 2.66E-08 3.12E-08 4.10E-08 6.15E-08 1.08E-03 1.427E-04 2.03E-08 Ag-110m 3.46E-10 2.54E-10 4.23E-10 1.90E-10 3.11E-10 2.44E-10 4.55E-09 2.57E-05 1.731E-05 3.00E-10 Sb-125 7.22E-09 2.34E-08 1.48E-08 1.59E-08 1.56E-08 2.74E-08 1.67E-07 1.10E-03 6.138E-04 3.77E-07 I-129 1.97E-10 1.82E-09 2.00E-10 2.02E-09 1.81E-09 2.28E-09 6.42E-10 3.65E-05 9.688E-06 9.39E-11 Cs-134 4.10E-08 3.01E-08 4.86E-08 4.38E-08 3.70E-08 3.66E-08 4.17E-07 2.66E-03 1.537E-03 2.36E-06 Cs-137 1.29E-06 1.20E-05 1.31E-06 1.32E-05 1.19E-05 1.50E-05 4.21E-06 2.40E-01 6.364E-02 4.05E-03 Ce-144 3.34E-10 9.49E-10 9.34E-10 5.80E-10 6.79E-10 1.25E-09 1.43E-08 7.73E-05 5.500E-05 3.03E-09 Pm-147 1.03E-11 5.60E-10 5.33E-11 5.60E-11 5.24E-11 1.80E-10 1.73E-09 1.08E-05 6.750E-06 4.56E-11 Eu-154 2.11E-09 2.99E-09 7.09E-09 1.30E-09 2.73E-09 3.62E-09 1.61E-08 1.46E-04 5.560E-05 3.09E-09 Eu-155 2.31E-10 7.27E-10 7.81E-10 4.29E-10 4.97E-10 8.69E-10 8.82E-10 1.80E-05 2.671E-06 7.14E-12 Pu-238 1.20E-10 2.14E-09 2.46E-09 7.60E-10 2.05E-09 1.45E-09 3.21E-08 1.67E-04 1.248E-04 1.56E-08 Pu-239 6.56E-11 8.10E-10 1.20E-09 4.20E-10 1.08E-09 7.08E-10 1.35E-08 7.22E-05 5.203E-05 2.71E-09 Pu-240 6.56E-11 8.10E-10 1.20E-09 4.20E-10 1.08E-09 7.08E-10 1.35E-08 7.22E-05 5.201E-05 2.71E-09 Pu-241 1.83E-10 9.92E-09 9.44E-10 9.92E-10 9.28E-10 3.18E-09 3.07E-08 1.91E-04 1.196E-04 1.43E-08 Am-241 8.24E-11 1.62E-09 1.07E-09 4.15E-10 1.51E-09 1.04E-09 2.70E-08 1.33E-04 1.059E-04 1.12E-08 Cm-242 4.44E-16 8.58E-15 6.11E-15 2.45E-15 9.23E-15 4.45E-15 1.84E-13 8.74E-10 7.261E-10 5.27E-19 Cm-243 5.58E-12 8.84E-11 7.47E-11 2.79E-11 1.02E-10 5.74E-11 2.00E-09 9.57E-06 7.887E-06 6.22E-11 Cm-244 4.21E-12 6.67E-11 5.64E-11 2.11E-11 7.70E-11 4.33E-11 1.51E-09 7.22E-06 5.953E-06 3.54E-11 Mean

• 1 dpm average value times 18,000/0.6324 (obs average beta fraction) = 2.846E+04 2.80E-01 Mean Dose from non-detectable nuclides: 5.07E-03 Standard Deviation of the Mean Mean Dose from detectable (Bolded) nuclides: 2.75E-01 6.59E-02

Attachment 2F Table 3 Analysis of Concrete Sample Variance Nuclide Fractions and Dose After Removal of MDA Nuclides Page 13 of 19 Column # ==> 2 3 4 5 6 7 8 9 10 11 12 13 23 PAB 11' West Dose For Fuel Bldg Dose For Dose For Dose For Pipe Trench 1.80E+04 Decon Room 1.80E+04 Spray Bldg 11' 1.80E+04 RCA Bldg 21' 1.80E+04 1FL1 dpm/100 cm2 01FL31 dpm/100 cm2 01FL41 dpm/100 cm2 01FL61 dpm/100 cm2 2004 1 dpm dose detectable 2004 1 dpm dose detectable 2004 1 dpm dose detectable 2004 1 dpm dose detectable 1.0 dpm Nuclide nf Times nf 2004 nf Times nf 2004 nf Times nf 2004 nf Times nf 2004 dose factor H-3 4.099E-04 1.373E-08 1.021E-03 6.510E-04 2.181E-08 5.020E-04 1.271E-03 4.260E-08 6.313E-03 1.049E-03 3.514E-08 7.230E-04 3.351E-05 Fe-55 2.928E-03 1.711E-09 1.271E-04 4.750E-04 2.775E-10 6.387E-06 3.988E-03 2.330E-09 3.453E-04 2.963E-03 1.731E-09 3.563E-05 5.843E-07 Co-57 2.934E-05 7.443E-12 5.531E-07 9.373E-05 2.378E-11 5.472E-07 1.042E-04 2.644E-11 3.918E-06 5.851E-05 1.484E-11 3.054E-07 2.537E-07 Co-60 1.578E-01 9.948E-07 7.393E-02 4.216E-03 2.658E-08 6.118E-04 3.280E-02 2.068E-07 3.065E-02 2.754E-02 1.736E-07 3.573E-03 6.305E-06 Ni-63 7.544E-01 8.428E-08 6.263E-03 2.167E-01 2.420E-08 5.571E-04 8.732E-01 9.755E-08 1.446E-02 1.212E-01 1.354E-08 2.786E-04 1.117E-07 Sr-90 1.349E-04 8.514E-09 6.327E-04 1.245E-03 7.857E-08 1.808E-03 5.690E-04 3.590E-08 5.320E-03 1.296E-03 8.180E-08 1.683E-03 6.310E-05 Cs-134 1.763E-03 4.205E-08 3.125E-03 1.313E-03 3.131E-08 7.206E-04 2.140E-03 5.102E-08 7.561E-03 1.877E-03 4.476E-08 9.211E-04 2.384E-05 Cs-137 8.254E-02 1.325E-06 9.847E-02 7.753E-01 1.245E-05 2.865E-01 8.595E-02 1.380E-06 2.045E-01 8.440E-01 1.355E-05 2.788E-01 1.605E-05 sum 1.000E+00 1.000E+00 1.000E+00 1.000E+00 obs. fraction 2.422E-01 7.821E-01 1.215E-01 8.748E-01 avg. fraction 5.051E-01 Nuclide Fractions and Dose After Removal of MDA Nuclides (Continued from above)

Column # ==> 14 15 16 17 18 19 20 21 22 23 PAB 11' Dose For CTMT Dose For CTMT Dose For Pipe Trench 1.80E+04 -2' Loop 2 1.80E+04 -2' Loop 1 1.80E+04 01FL81 dpm/100 cm2 02FL21 dpm/100 cm2 02FL51 dpm/100 cm2 2004 1 dpm dose detectable 2004 1 dpm dose detectable 2004 1 dpm dose detectable 1.0 dpm Nuclide nf Times nf 2004 nf Times nf 2004 nf Times nf 2004 dose factor H-3 3.685E-03 1.235E-07 2.726E-03 4.450E-03 1.491E-07 2.742E-03 1.533E-01 5.138E-06 1.850E-01 3.351E-05 Fe-55 2.440E-03 1.425E-09 3.147E-05 5.443E-04 3.180E-10 5.848E-06 2.079E-02 1.215E-08 4.374E-04 5.843E-07 Co-57 6.662E-05 1.690E-11 3.731E-07 1.203E-04 3.053E-11 5.614E-07 1.780E-03 4.515E-10 1.625E-05 2.537E-07 Co-60 5.193E-02 3.274E-07 7.228E-03 6.228E-03 3.927E-08 7.221E-04 1.307E-01 8.240E-07 2.967E-02 6.305E-06 Ni-63 1.785E-01 1.994E-08 4.403E-04 1.605E-02 1.793E-09 3.298E-05 3.241E-01 3.621E-08 1.304E-03 1.117E-07 Sr-90 4.508E-04 2.844E-08 6.280E-04 3.997E-04 2.522E-08 4.638E-04 1.550E-02 9.779E-07 3.521E-02 6.310E-05 Cs-134 1.596E-03 3.806E-08 8.404E-04 1.595E-03 3.803E-08 6.993E-04 2.211E-02 5.271E-07 1.898E-02 2.384E-05 Cs-137 7.613E-01 1.222E-05 2.699E-01 9.706E-01 1.558E-05 2.866E-01 3.317E-01 5.325E-06 1.917E-01 1.605E-05 sum 1.000E+00 1.000E+00 1.000E+00 obs. fraction 8.153E-01 9.788E-01 5.000E-01

Table 4 Attachment 2F Mean and Standard Deviation of Dose Analysis of Concrete Sample Variance Using the Average of Fractions and Individual Core Methods Page 14 of 19 Column # ==> 1 2 3 4 5 6 7 8 9 10 11 7 Cores Dose Results (in mrem) for Average of the Fractions (1.0 dpm)

Nuclide Mean nf 1FL1 01FL31 01FL41 01FL61 01FL81 02FL21 02FL51 mean* stdev mean (stdev Mean)2 H-3 2.36E-02 1.37E-08 2.18E-08 4.26E-08 3.51E-08 1.23E-07 1.49E-07 5.14E-06 2.30E-02 2.117E-02 4.48E-04 Fe-55 4.81E-03 1.71E-09 2.78E-10 2.33E-09 1.73E-09 1.43E-09 3.18E-10 1.21E-08 8.32E-05 4.602E-05 2.12E-09 Co-57 3.06E-04 7.44E-12 2.38E-11 2.64E-11 1.48E-11 1.69E-11 3.05E-11 4.51E-10 2.38E-06 1.802E-06 3.25E-12 Co-60 5.84E-02 9.95E-07 2.66E-08 2.07E-07 1.74E-07 3.27E-07 3.93E-08 8.24E-07 1.08E-02 4.253E-03 1.81E-05 Ni-63 3.55E-01 8.43E-08 2.42E-08 9.76E-08 1.35E-08 1.99E-08 1.79E-09 3.62E-08 1.16E-03 4.055E-04 1.64E-07 Sr-90 2.80E-03 8.51E-09 7.86E-08 3.59E-08 8.18E-08 2.84E-08 2.52E-08 9.78E-07 5.16E-03 3.912E-03 1.53E-05 Cs-134 4.56E-03 4.20E-08 3.13E-08 5.10E-08 4.48E-08 3.81E-08 3.80E-08 5.27E-07 3.22E-03 2.029E-03 4.12E-06 Cs-137 5.50E-01 1.33E-06 1.24E-05 1.38E-06 1.36E-05 1.22E-05 1.56E-05 5.32E-06 2.58E-01 6.632E-02 4.40E-03 Mean

• 1 dpm average value times 18,000/0.6164 (obs average beta fraction) = 2.920E+04 3.01E-01 Standard Deviation Of the Mean 6.99E-02 7 Core Dose Results From Individual Cores 18,000 dpm/100 cm2 Observable Beta 01FL1 01FL31 01FL41 01FL61 01FL81 02FL21 02FL51 Nuclide 2004 2004 2004 2004 2004 2004 2004 H-3 1.02E-03 5.02E-04 6.31E-03 7.23E-04 2.73E-03 2.74E-03 1.85E-01 Fe-55 1.27E-04 6.39E-06 3.45E-04 3.56E-05 3.15E-05 5.85E-06 4.37E-04 Co-57 5.53E-07 5.47E-07 3.92E-06 3.05E-07 3.73E-07 5.61E-07 1.63E-05 Co-60 7.39E-02 6.12E-04 3.06E-02 3.57E-03 7.23E-03 7.22E-04 2.97E-02 Ni-63 6.26E-03 5.57E-04 1.45E-02 2.79E-04 4.40E-04 3.30E-05 1.30E-03 Sr-90 6.33E-04 1.81E-03 5.32E-03 1.68E-03 6.28E-04 4.64E-04 3.52E-02 Cs-134 3.12E-03 7.21E-04 7.56E-03 9.21E-04 8.40E-04 6.99E-04 1.90E-02 Cs-137 9.85E-02 2.86E-01 2.05E-01 2.79E-01 2.70E-01 2.87E-01 1.92E-01 dose sum 1.84E-01 2.91E-01 2.69E-01 2.86E-01 2.82E-01 2.91E-01 4.62E-01 Individual Core (7) Propagation of Error Mean 2.95E-01 Standard Deviation of the Mean 3.14E-02

Table 5 Attachment 2F Nuclide Fractions and Dose From Eight Additional Cores Analysis of Concrete Sample Variance (Table 5 page1 of 2) Page 15 of 19 Column # ==> 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 PAB Containment Containment Containment Evaporator Loop 2 Dose For Loop 1 Dose For Loop 3 Dose For Cubicle Dose For CA9900 1.80E+04 CA9900 1.80E+04 CA9900 1.80E+04 CA9900 1.80E+04 12-C003-A dpm/100 cm2 12-C004-A dpm/100 cm2 12-C005-A dpm/100 cm2 13-C001-A dpm/100 cm2 2004 1 dpm dose detectable 2004 1 dpm dose detectable 2004 1 dpm dose detectable 2004 1 dpm dose detectable Nuclide nf Times nf 2004 nf Times nf 2.00E+03 nf Times nf 2004 nf Times nf 2004 H-3 6.299E-03 2.11E-07 4.62E-03 3.601E-02 1.21E-06 2.58E-02 2.101E-03 7.038E-08 1.60E-03 7.211E-02 2.42E-06 5.54E-02 C-14 < 3.284E-04 5.92E-09 1.30E-04 < 8.893E-04 1.60E-08 3.42E-04 < 5.168E-05 9.318E-10 2.11E-05 < 3.268E-02 5.89E-07 1.35E-02 Mn-54 1.329E-05 2.80E-11 6.12E-07 2.928E-05 6.16E-11 1.32E-06 < 5.172E-05 1.088E-10 2.47E-06 < 2.222E-04 4.67E-10 1.07E-05 Fe-55 3.593E-02 2.10E-08 4.60E-04 8.846E-03 5.17E-09 1.10E-04 1.188E-03 6.941E-10 1.58E-05 8.482E-03 4.96E-09 1.14E-04 Co-57 < 2.762E-05 7.01E-12 1.53E-07 < 4.709E-05 1.19E-11 2.55E-07 < 2.070E-04 5.250E-11 1.19E-06 < 7.788E-04 1.98E-10 4.53E-06 Co-58 < 3.161E-09 2.56E-15 5.59E-11 < 3.862E-09 3.12E-15 6.67E-11 < 2.284E-08 1.847E-14 4.19E-10 < 8.250E-08 6.67E-14 1.53E-09 Ni-59 < 1.359E-03 1.64E-11 3.60E-07 < 1.122E-03 1.36E-11 2.90E-07 < 2.046E-03 2.473E-11 5.61E-07 < 1.234E-03 1.49E-11 3.42E-07 Co-60 8.522E-02 5.37E-07 1.18E-02 4.633E-02 2.92E-07 6.24E-03 4.069E-02 2.566E-07 5.82E-03 8.315E-02 5.24E-07 1.20E-02 Ni-63 1.330E-01 1.49E-08 3.25E-04 1.098E-01 1.23E-08 2.62E-04 2.002E-01 2.237E-08 5.08E-04 1.208E-01 1.35E-08 3.10E-04 Zn-65 < 6.124E-06 6.48E-11 1.42E-06 < 1.277E-05 1.35E-10 2.88E-06 < 6.208E-05 6.566E-10 1.49E-05 < 2.748E-04 2.91E-09 6.67E-05 Sr-90 9.839E-04 6.21E-08 1.36E-03 1.330E-03 8.39E-08 1.79E-03 1.131E-03 7.135E-08 1.62E-03 < 4.323E-03 2.73E-07 6.26E-03 Nb-94 < 5.375E-05 9.00E-11 1.97E-06 < 6.424E-05 1.08E-10 2.30E-06 < 4.061E-04 6.798E-10 1.54E-05 < 1.471E-03 2.46E-09 5.65E-05 Tc-99 < 1.338E-04 4.30E-08 9.42E-04 < 1.443E-04 4.64E-08 9.91E-04 < 1.363E-04 4.382E-08 9.94E-04 < 1.177E-04 3.79E-08 8.68E-04 Ru-106 < 9.500E-05 1.17E-09 2.56E-05 < 1.483E-04 1.83E-09 3.90E-05 < 7.091E-04 8.733E-09 1.98E-04 < 2.747E-03 3.38E-08 7.76E-04 Ag-110m < 1.072E-04 3.70E-10 8.09E-06 < 1.847E-04 6.37E-10 1.36E-05 < 4.383E-05 1.512E-10 3.43E-06 < 2.733E-03 9.43E-09 2.16E-04 Sb-125 2.306E-04 4.63E-10 1.01E-05 < 2.742E-04 5.50E-10 1.18E-05 < 1.229E-03 2.467E-09 5.60E-05 < 4.156E-03 8.34E-09 1.91E-04 I-129 < 2.673E-07 1.73E-09 3.80E-05 < 2.876E-07 1.87E-09 3.99E-05 < 2.718E-07 1.763E-09 4.00E-05 < 2.354E-07 1.53E-09 3.50E-05 Cs-134 1.658E-03 3.95E-08 8.65E-04 2.244E-03 5.35E-08 1.14E-03 1.196E-03 2.851E-08 6.47E-04 < 8.118E-04 1.94E-08 4.44E-04 Cs-137 7.332E-01 1.18E-05 2.58E-01 7.910E-01 1.27E-05 2.71E-01 7.465E-01 1.198E-05 2.72E-01 6.465E-01 1.04E-05 2.38E-01 Ce-144 < 3.068E-05 3.18E-11 6.96E-07 < 5.077E-05 5.26E-11 1.12E-06 < 2.371E-04 2.457E-10 5.58E-06 < 8.396E-04 8.70E-10 2.00E-05 Pm-147 < 9.913E-05 2.01E-10 4.40E-06 < 6.856E-05 1.39E-10 2.97E-06 < 1.939E-05 3.933E-11 8.92E-07 < 7.612E-04 1.54E-09 3.54E-05 Eu-154 < 1.007E-04 7.75E-11 1.70E-06 < 1.202E-04 9.24E-11 1.97E-06 < 7.820E-04 6.014E-10 1.36E-05 < 3.462E-03 2.66E-09 6.11E-05 Eu-155 < 1.068E-04 1.05E-11 2.30E-07 < 1.767E-04 1.74E-11 3.71E-07 < 7.920E-04 7.784E-11 1.77E-06 < 3.559E-03 3.50E-10 8.03E-06 Pu-238 < 1.049E-05 1.46E-09 3.20E-05 < 1.013E-05 1.41E-09 3.01E-05 < 1.550E-06 2.157E-10 4.89E-06 < 1.994E-04 2.78E-08 6.37E-04 Pu-239 < 6.719E-06 1.03E-09 2.26E-05 < 4.201E-06 6.47E-10 1.38E-05 < 1.052E-06 1.620E-10 3.68E-06 < 8.581E-05 1.32E-08 3.03E-04 Pu-240 < 6.718E-06 1.03E-09 2.26E-05 < 4.200E-06 6.47E-10 1.38E-05 < 1.052E-06 1.620E-10 3.68E-06 < 8.579E-05 1.32E-08 3.03E-04 Pu-241 < 8.625E-04 2.57E-09 5.64E-05 < 5.965E-04 1.78E-09 3.80E-05 < 1.687E-04 5.036E-10 1.14E-05 < 6.623E-03 1.98E-08 4.54E-04 Am-241 < 9.807E-05 4.57E-09 1.00E-04 < 3.344E-04 1.56E-08 3.33E-04 < 2.645E-05 1.232E-09 2.80E-05 < 1.383E-03 6.44E-08 1.48E-03 Cm-242 < 5.750E-07 4.03E-13 8.81E-09 < 1.944E-06 1.36E-12 2.91E-08 < 1.638E-07 1.147E-13 2.60E-09 < 8.893E-06 6.23E-12 1.43E-07 Cm-243 < 1.801E-05 2.76E-10 6.04E-06 < 6.190E-05 9.48E-10 2.02E-05 < 4.282E-06 6.555E-11 1.49E-06 < 2.123E-04 3.25E-09 7.46E-05 Cm-244 < 1.725E-05 2.13E-10 4.65E-06 < 5.929E-05 7.30E-10 1.56E-05 < 4.102E-06 5.053E-11 1.15E-06 < 2.034E-04 2.51E-09 5.75E-05 sum 1.000E+00 1.27E-05 2.78E-01 1.000E+00 1.44E-05 3.08E-01 1.000E+00 1.25E-05 2.84E-01 1.000E+00 1.45E-05 3.32E-01 obs. fraction 8.222E-01 8.429E-01 7.932E-01 7.845E-01

Table 5 Attachment 2F Nuclide Fractions and Dose From Eight Additional Cores Analysis of Concrete Variance (Table 5 page 2 of 2) Page 16 of 19 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 PAB PAB Pipe Tunnel Dose For Pipe Tunnel Dose For O/A Trench Dose For O/A Trench Dose For CA9900 1.80E+04 CA9900 1.80E+04 CA9900 1.80E+04 CA9900 1.80E+04 13-C002-A dpm/100 cm2 13-C003-A dpm/100 cm2 12-C001-A dpm/100 cm2 12-C002-A dpm/100 cm2 2004 1 dpm dose detectable 2004 1 dpm dose detectable 2004 1 dpm dose detectable 2004 1 dpm dose detectable 1.0 dpm Nuclide nf Times nf 2004 nf Times nf 2004 nf Times nf 2004 nf Times nf 2004 dose factor H-3 < 3.936E-03 1.319E-07 7.02E-03 < 2.169E-02 7.269E-07 2.88E-02 < 5.168E-03 1.732E-07 3.29E-03 < 4.438E-03 1.487E-07 2.88E-03 3.351E-05 C-14 < 3.618E-03 6.525E-08 3.48E-03 < 1.058E-02 1.908E-07 7.56E-03 < 3.978E-02 7.174E-07 1.36E-02 < 2.851E-02 5.141E-07 9.95E-03 1.803E-05 Mn-54 < 1.579E-04 3.321E-10 1.77E-05 < 9.359E-04 1.969E-09 7.80E-05 3.607E-04 7.587E-10 1.44E-05 4.206E-05 8.848E-11 1.71E-06 2.104E-06 Fe-55 2.348E-02 1.372E-08 7.31E-04 5.639E-02 3.295E-08 1.31E-03 1.015E-03 5.929E-10 1.13E-05 2.553E-03 1.492E-09 2.89E-05 5.843E-07 Co-57 < 3.182E-04 8.073E-11 4.30E-06 < 1.881E-03 4.772E-10 1.89E-05 < 3.853E-04 9.774E-11 1.86E-06 < 4.736E-04 1.202E-10 2.33E-06 2.537E-07 Co-58 < 5.662E-08 4.578E-14 2.44E-09 < 3.615E-07 2.923E-13 1.16E-08 < 4.889E-08 3.953E-14 7.52E-10 < 6.306E-08 5.099E-14 9.87E-10 8.086E-07 Ni-59 < 6.317E-03 7.636E-11 4.07E-06 < 4.415E-03 5.337E-11 2.11E-06 < 4.305E-04 5.204E-12 9.90E-08 < 5.810E-04 7.023E-12 1.36E-07 1.209E-08 Co-60 9.653E-02 6.086E-07 3.24E-02 2.071E-01 1.306E-06 5.17E-02 5.186E-01 3.270E-06 6.22E-02 5.516E-01 3.478E-06 6.73E-02 6.305E-06 Ni-63 6.183E-01 6.908E-08 3.68E-03 4.322E-01 4.828E-08 1.91E-03 4.215E-02 4.708E-09 8.96E-05 5.687E-02 6.354E-09 1.23E-04 1.117E-07 Zn-65 < 2.096E-04 2.216E-09 1.18E-04 < 1.271E-03 1.344E-08 5.32E-04 < 3.116E-04 3.295E-09 6.27E-05 < 4.043E-04 4.276E-09 8.27E-05 1.058E-05 Sr-90 < 2.305E-03 1.455E-07 7.75E-03 1.651E-02 1.042E-06 4.13E-02 3.599E-03 2.271E-07 4.32E-03 3.117E-03 1.967E-07 3.81E-03 6.310E-05 Nb-94 < 1.065E-03 1.782E-09 9.49E-05 < 7.745E-03 1.296E-08 5.14E-04 < 1.990E-03 3.331E-09 6.33E-05 < 2.535E-03 4.243E-09 8.21E-05 1.674E-06 Tc-99 < 4.053E-05 1.303E-08 6.94E-04 < 3.150E-05 1.013E-08 4.01E-04 < 6.755E-05 2.172E-08 4.13E-04 < 6.028E-05 1.939E-08 3.75E-04 3.216E-04 Ru-106 < 1.619E-03 1.993E-08 1.06E-03 < 8.987E-03 1.107E-07 4.38E-03 < 2.258E-03 2.781E-08 5.29E-04 < 2.832E-03 3.487E-08 6.75E-04 1.232E-05 Ag-110m < 1.256E-03 4.333E-09 2.31E-04 < 6.093E-03 2.102E-08 8.33E-04 < 1.708E-03 5.890E-09 1.12E-04 < 2.147E-03 7.404E-09 1.43E-04 3.449E-06 Sb-125 < 2.264E-03 4.544E-09 2.42E-04 < 8.713E-03 1.749E-08 6.93E-04 2.827E-03 5.674E-09 1.08E-04 < 3.093E-03 6.208E-09 1.20E-04 2.007E-06 I-129 < 8.091E-08 5.250E-10 2.80E-05 < 6.277E-08 4.073E-10 1.61E-05 < 1.346E-07 8.733E-10 1.66E-05 < 1.201E-07 7.791E-10 1.51E-05 6.489E-03 Cs-134 4.255E-03 1.014E-07 5.40E-03 < 3.114E-03 7.424E-08 2.94E-03 1.951E-03 4.651E-08 8.85E-04 1.361E-03 3.246E-08 6.28E-04 2.384E-05 Cs-137 2.222E-01 3.567E-06 1.90E-01 1.725E-01 2.769E-06 1.10E-01 3.701E-01 5.942E-06 1.13E-01 3.303E-01 5.303E-06 1.03E-01 1.605E-05 Ce-144 < 3.600E-04 3.731E-10 1.99E-05 < 2.155E-03 2.233E-09 8.85E-05 < 4.181E-04 4.333E-10 8.24E-06 < 5.199E-04 5.389E-10 1.04E-05 1.036E-06 Pm-147 < 7.735E-04 1.569E-09 8.36E-05 < 1.766E-03 3.581E-09 1.42E-04 < 1.674E-04 3.395E-10 6.46E-06 < 1.755E-04 3.560E-10 6.89E-06 2.028E-06 Eu-154 < 2.060E-03 1.584E-09 8.44E-05 < 9.923E-03 7.631E-09 3.02E-04 < 3.405E-03 2.619E-09 4.98E-05 < 4.419E-03 3.398E-09 6.58E-05 7.690E-07 Eu-155 < 1.245E-03 1.224E-10 6.52E-06 < 8.223E-03 8.081E-10 3.20E-05 < 1.787E-03 1.756E-10 3.34E-06 < 2.288E-03 2.249E-10 4.35E-06 9.828E-08 Pu-238 < 6.927E-05 9.641E-09 5.14E-04 2.898E-04 4.033E-08 1.60E-03 3.849E-05 5.357E-09 1.02E-04 3.581E-05 4.984E-09 9.64E-05 1.392E-04 Pu-239 < 2.271E-05 3.497E-09 1.86E-04 2.607E-04 4.015E-08 1.59E-03 1.411E-05 2.172E-09 4.13E-05 2.565E-05 3.949E-09 7.64E-05 1.540E-04 Pu-240 < 2.270E-05 3.496E-09 1.86E-04 2.607E-04 4.014E-08 1.59E-03 1.411E-05 2.172E-09 4.13E-05 2.564E-05 3.948E-09 7.64E-05 1.540E-04 Pu-241 < 6.730E-03 2.009E-08 1.07E-03 1.536E-02 4.584E-08 1.82E-03 1.457E-03 4.348E-09 8.27E-05 1.527E-03 4.557E-09 8.82E-05 2.985E-06 Am-241 < 6.356E-04 2.961E-08 1.58E-03 1.557E-03 7.252E-08 2.87E-03 1.681E-05 7.832E-10 1.49E-05 < 2.242E-06 1.045E-10 2.02E-06 4.658E-05 Cm-242 < 3.505E-06 2.454E-12 1.31E-07 < 9.076E-07 6.356E-13 2.52E-08 < 4.845E-09 3.393E-15 6.45E-11 < 4.715E-09 3.302E-15 6.39E-11 7.002E-07 Cm-243 < 1.086E-04 1.663E-09 8.86E-05 < 6.707E-05 1.027E-09 4.07E-05 1.208E-06 1.850E-11 3.52E-07 < 2.123E-07 3.251E-12 6.29E-08 1.531E-05 Cm-244 < 1.041E-04 1.282E-09 6.83E-05 < 6.425E-05 7.914E-10 3.14E-05 1.158E-06 1.426E-11 2.71E-07 < 2.034E-07 2.505E-12 4.85E-08 1.232E-05 sum 1.000E+00 2.57E-01 1.000E+00 6.63E-06 2.63E-01 1.000E+00 1.05E-05 1.99E-01 1.000E+00 9.78E-06 1.89E-01 obs. fraction 3.379E-01 4.544E-01 9.464E-01 9.302E-01

Table 6 Attachment 2F Nuclide Fraction and Dose After Removal of MDA Nuclides Analysis of Concrete Variance O/A Trench and PAB Pipe Tunnel Page 17 of 19 (Table 6 Page 1 of 2)

Column # ==> 1 2 3 4 5 6 7 8 9 10 7/30/02 Pipe Tunnel Pipe Tunnel Dose For Pipe Tunnel Pipe Tunnel Dose For CA9900 CA9900 1.80E+04 CA9900 CA9900 1.80E+04 13-C002-A 13-C002-A dpm/100 cm2 13-C003-A 13-C003-A dpm/100 cm2 2004 2004 1 dpm dose detectable 2004 2004 1 dpm dose detectable Nuclide initial nf normalized nf Times nf 2004 initial nf normalized nf Times nf 2004 Mn-54 < 1.579E-04 1.616E-04 3.40E-10 1.83E-05 < 9.359E-04 1.023E-03 2.151E-09 8.689E-05 Fe-55 2.348E-02 2.403E-02 1.40E-08 7.54E-04 5.639E-02 6.162E-02 3.600E-08 1.454E-03 Co-60 9.653E-02 9.878E-02 6.23E-07 3.34E-02 2.071E-01 2.262E-01 1.426E-06 5.762E-02 Ni-63 6.183E-01 6.328E-01 7.07E-08 3.80E-03 4.322E-01 4.722E-01 5.275E-08 2.131E-03 Sr-90 < 2.305E-03 2.359E-03 1.49E-07 7.99E-03 1.651E-02 1.804E-02 1.138E-06 4.597E-02 Sb-125 < 2.264E-03 2.317E-03 4.65E-09 2.50E-04 < 8.713E-03 9.520E-03 1.911E-08 7.718E-04 Cs-134 4.255E-03 4.354E-03 1.04E-07 5.58E-03 < 3.114E-03 3.402E-03 8.112E-08 3.276E-03 Cs-137 2.222E-01 2.274E-01 3.65E-06 1.96E-01 1.725E-01 1.884E-01 3.025E-06 1.222E-01 Pu-238 < 6.927E-05 7.088E-05 9.87E-09 5.30E-04 2.898E-04 3.166E-04 4.407E-08 1.780E-03 Pu-239 < 2.271E-05 2.324E-05 3.58E-09 1.92E-04 2.607E-04 2.849E-04 4.386E-08 1.772E-03 Pu-240 < 2.270E-05 2.323E-05 3.58E-09 1.92E-04 2.607E-04 2.848E-04 4.385E-08 1.771E-03 Pu-241 < 6.730E-03 6.888E-03 2.06E-08 1.10E-03 1.536E-02 1.678E-02 5.009E-08 2.023E-03 Am-241 < 6.356E-04 6.505E-04 3.03E-08 1.63E-03 1.557E-03 1.701E-03 7.924E-08 3.201E-03 Cm-243 < 1.086E-04 1.112E-04 1.70E-09 9.14E-05 < 6.707E-05 7.328E-05 1.122E-09 4.531E-05 Cm-244 < 1.041E-04 1.065E-04 1.31E-09 7.05E-05 < 6.425E-05 7.020E-05 8.647E-10 3.493E-05 sum 9.772E-01 1.000E+00 4.686E-06 2.556E-01 9.152E-01 1.000E+00 6.061E-06 2.517E-01 obs. b fraction 3.28E-01 3.35E-01 4.08E-01 4.46E-01

Table 6 Attachment 2F Nuclide Fractions and Dose After Removal of MDA Nuclides Analysis of Concrete Variance O/A Trench and PAB Pipe Tunnel Page 18 of 19 (Table 6 page 2 of 2)

Column # ==> 11 12 13 14 15 16 17 18 19 20 O/A Trench O/A Trench Dose For O/A Trench O/A Trench Dose For CA9900 CA9900 1.80E+04 CA9900 CA9900 1.80E+04 12-C001-A 12-C001-A dpm/100 cm2 12-C002-A 12-C002-A dpm/100 cm2 2004 2004 1 dpm dose detectable 2004 2004 1 dpm dose detectable Nuclide initial nf normalized nf Times nf 2004 initial nf normalized nf Times nf 2004 Mn-54 3.607E-04 3.828E-04 8.053E-10 1.522E-05 4.206E-05 4.425E-05 9.308E-11 1.790E-06 Fe-55 1.015E-03 1.077E-03 6.293E-10 1.190E-05 2.553E-03 2.685E-03 1.569E-09 3.018E-05 Co-60 5.186E-01 5.504E-01 3.470E-06 6.561E-02 5.516E-01 5.803E-01 3.659E-06 7.038E-02 Ni-63 4.215E-02 4.473E-02 4.998E-09 9.448E-05 5.687E-02 5.982E-02 6.684E-09 1.286E-04 Sr-90 3.599E-03 3.820E-03 2.411E-07 4.557E-03 3.117E-03 3.279E-03 2.069E-07 3.980E-03 Sb-125 2.827E-03 3.000E-03 6.022E-09 1.138E-04 < 3.093E-03 3.253E-03 6.530E-09 1.256E-04 Cs-134 1.951E-03 2.071E-03 4.937E-08 9.333E-04 1.361E-03 1.432E-03 3.415E-08 6.568E-04 Cs-137 3.701E-01 3.928E-01 6.307E-06 1.192E-01 3.303E-01 3.475E-01 5.579E-06 1.073E-01 Pu-238 3.849E-05 4.085E-05 5.686E-09 1.075E-04 3.581E-05 3.767E-05 5.243E-09 1.008E-04 Pu-239 1.411E-05 1.498E-05 2.306E-09 4.359E-05 2.565E-05 2.698E-05 4.154E-09 7.991E-05 Pu-240 1.411E-05 1.497E-05 2.305E-09 4.358E-05 2.564E-05 2.697E-05 4.153E-09 7.989E-05 Pu-241 1.457E-03 1.546E-03 4.615E-09 8.724E-05 1.527E-03 1.606E-03 4.794E-09 9.222E-05 Am-241 1.681E-05 1.785E-05 8.313E-10 1.571E-05 < 2.242E-06 2.359E-06 1.099E-10 2.114E-06 Cm-243 1.208E-06 1.283E-06 1.963E-11 3.712E-07 < 2.123E-07 2.234E-07 3.419E-12 6.577E-08 Cm-244 1.158E-06 1.229E-06 1.513E-11 2.861E-07 < 2.034E-07 2.140E-07 2.636E-12 5.070E-08 sum 9.421E-01 1.000E+00 1.010E-05 1.909E-01 9.506E-01 1.000E+00 9.512E-06 1.830E-01 obs. b fraction 8.97E-01 9.52E-01 8.90E-01 9.36E-01

Table 7 Attachment 2F Mean and Standard Deviation of Dose Analysis of Concrete Sample Variance For TRU Affected Cores Page 19 of 19 Using the Average of the Fractions and Individual Core Method Column # ==> 2 3 4 5 6 7 8 9 4 Cores Dose Results (in mrem) for Average of the Fractions (1.0 dpm)

Nuclide Mean nf 13-C002-A 13-C003-A 12-C001-A 12-C002-A mean* stdev mean (stdev Mean)2 Mn-54 4.028E-04 3.40E-10 2.151E-09 8.053E-10 9.308E-11 2.286E-05 1.238E-05 1.533E-10 Fe-55 2.235E-02 1.40E-08 3.600E-08 6.293E-10 1.569E-09 3.523E-04 2.222E-04 4.935E-08 Co-60 3.639E-01 6.23E-07 1.426E-06 3.470E-06 3.659E-06 6.190E-02 2.030E-02 4.120E-04 Ni-63 3.024E-01 7.07E-08 5.275E-08 4.998E-09 6.684E-09 9.114E-04 4.464E-04 1.993E-07 Sr-90 6.874E-03 1.49E-07 1.138E-06 2.411E-07 2.069E-07 1.170E-02 6.355E-03 4.039E-05 Sb-125 4.523E-03 4.65E-09 1.911E-08 6.022E-09 6.530E-09 2.449E-04 9.084E-05 8.252E-09 Cs-134 2.815E-03 1.04E-07 8.112E-08 4.937E-08 3.415E-08 1.811E-03 4.226E-04 1.786E-07 Cs-137 2.890E-01 3.65E-06 3.025E-06 6.307E-06 5.579E-06 1.252E-01 2.097E-02 4.395E-04 Pu-238 1.165E-04 9.87E-09 4.407E-08 5.686E-09 5.243E-09 4.375E-04 2.520E-04 6.352E-08 Pu-239 8.752E-05 3.58E-09 4.386E-08 2.306E-09 4.154E-09 3.636E-04 2.735E-04 7.479E-08 Pu-240 8.750E-05 3.58E-09 4.385E-08 2.305E-09 4.153E-09 3.635E-04 2.734E-04 7.476E-08 Pu-241 6.705E-03 2.06E-08 5.009E-08 4.615E-09 4.794E-09 5.399E-04 2.886E-04 8.330E-08 Am-241 5.929E-04 3.03E-08 7.924E-08 8.313E-10 1.099E-10 7.452E-04 5.015E-04 2.515E-07 Cm-243 4.649E-05 1.70E-09 1.122E-09 1.963E-11 3.419E-12 1.920E-05 1.137E-05 1.292E-10 Cm-244 4.454E-05 1.31E-09 8.647E-10 1.513E-11 2.636E-12 1.480E-05 8.760E-06 7.674E-11 Mean

• 1 dpm average value times 18,000/0.6672 (obs. average beta fraction) = 2.698E+04 2.046E-01 Standard Deviation Of the Mean 2.988E-02 Core Dose Results From Individual Cores 18,000 dpm/100 cm2 Observable Beta Nuclide 13-C002-A 13-C003-A 12-C001-A 12-C002-A Mn-54 1.83E-05 8.689E-05 1.522E-05 1.790E-06 Fe-55 7.54E-04 1.454E-03 1.190E-05 3.018E-05 Co-60 3.34E-02 5.762E-02 6.561E-02 7.038E-02 Ni-63 3.80E-03 2.131E-03 9.448E-05 1.286E-04 Sr-90 7.99E-03 4.597E-02 4.557E-03 3.980E-03 Sb-125 2.50E-04 7.718E-04 1.138E-04 1.256E-04 Cs-134 5.58E-03 3.276E-03 9.333E-04 6.568E-04 Cs-137 1.96E-01 1.222E-01 1.192E-01 1.073E-01 Pu-238 5.30E-04 1.780E-03 1.075E-04 1.008E-04 Pu-239 1.92E-04 1.772E-03 4.359E-05 7.991E-05 Pu-240 1.92E-04 1.771E-03 4.358E-05 7.989E-05 Pu-241 1.10E-03 2.023E-03 8.724E-05 9.222E-05 Am-241 1.63E-03 3.201E-03 1.571E-05 2.114E-06 Cm-243 9.14E-05 4.531E-05 3.712E-07 6.577E-08 Cm-244 7.05E-05 3.493E-05 2.861E-07 5.070E-08 dose sum 2.517E-01 2.441E-01 1.909E-01 1.830E-01 Individual Core (4) Propagation of Error Mean 2.17E-01 Standard Deviation of the Mean 1.77E-02

MYAPC License Termination Plan Attachment 2G Revision 3 Page 1 of 3 October 15, 2002 ATTACHMENT 2G Supplemental Information Regarding Concrete Core Data Use

MYAPC License Termination Plan Attachment 2G Revision 3 Page 2 of 3 October 15, 2002 Supplemental Information Regarding Concrete Core Data Use To characterize contaminated concrete surfaces, there were three sets of concrete cores obtained and analyzed. The resulting data was used to establish the appropriate nuclide fractions and support the dose assessment in Section 6. Each core set was taken for different reasons and was analyzed by methods appropriate to each sets purpose. The following discussion summarizes the purpose of each set and key elements of the analysis for each.

A. Initial Set of Concrete Cores (Initial Site Characterization)

The first set of cores were collected during initial site characterization by GTS Duratek and were used to represent typical concrete nuclide data. Seven of these cores with the highest total activity were selected for off-site analysis to determine the amount of HTD nuclides present.

(Using the highest activity cores offered the best chance of detection for low activity HTD nuclides.) The HTDs were determined using radiochemical analytical techniques; gamma emitting nuclides were determined by gamma spectroscopy with the cores counted 21 inches above the detector to approximate a point source. The results from these cores formed the basis for the establishment of the contaminated concrete surface nuclide fraction for the majority of basement concrete surfaces (i.e., the balance of plant concrete surfaces). Certain subsequent core samples and analyses would lead to establishing a separate, unique nuclide fraction for limited areas warranting such treatment. This is discussed below. See Section 2.5.3a and F for additional detail.

B. Second Set of Concrete Cores Forty three (43) additional cores were collected during continuing site characterization. This data was used primarily for establishing Et for contaminated concrete. The number of cores obtained was established so that each building or plant area would have several cores included in the data analysis with the goal that the sample population, as a whole, would more accurately represent the nuclide ratios for concrete surfaces. These 43 core samples were processed for the Et determination by initially gross counting the cores, followed by gamma spectroscopy analysis.

The cores were counted initially onsite; six cores were later recounted at an offsite vendors laboratory (DES1). The onsite HPGe detectors had been calibrated using a concrete standard of uniform activity. The samples were counted at DES using a similar geometry, and the results showed good agreement. In order to determine total activity for the Et calculation, six of the cores were dissolved, and the dissolved material was again counted using the geometry specific to the analytical technique. The counting results for the dissolved cores showed that the activity was mostly on the surface of the concrete. (Later evaluation of the data using Microshield modeling verified that the Co-60 activity was located on the surface of the concrete and had a correction factor of approximately 0.5 while the Cs-137 activity was as deep as 1 mm in the core 1

Duke Engineering and Services Environmental Laboratory, now referred to as Framatome ANP DE&S Environmental Laboratory.

MYAPC License Termination Plan Attachment 2G Revision 3 Page 3 of 3 October 15, 2002 and had a correction factor of 0.73.) An average correction factor was determined to convert the activity from a surface count to the total activity in the core. The value of the correction factor was determined to be 0.68 from the DES data, as compared to 0.67 based on the onsite data.

C. Third Set of Concrete Cores Upon reviewing some of the original GTS data, a question remained concerning the possible presence of TRUs on concrete surfaces. A specific area of concern was the containment outer annulus trench. A decision was made to obtain cores on either side of an original trench sample to confirm or disprove the presence of TRUs. At the same time, additional cores were obtained to replace those destroyed by sample analysis. Three additional cores were collected from within the loops of containment, and three additional samples were collected in the PAB. Thus, the third set of cores totaled 8.

This set of core samples was analyzed by gross counting, gamma spectroscopy, and offsite analysis for HTD nuclides. This data formed the basis for the development of the alternate concrete nuclide fraction for trenches, pipe tunnel and other unique (special) areas, as discussed in Section 2.5.3.a and Attachment 2F. Background information related to the development of this nuclide fraction was described in a special report from the Technical Issue Resolution Process (TIRP). The report addressed a number of concerns related to the presence of TRUs in certain plant locations. (See Section 2.7, References.)

D. Core Data Adjustments The nuclide fraction given in Table 2-7 of the LTP was derived from the data provided by the seven original cores. Four of the additional eight cores were confirmed to be included in the balance of plant concrete surfaces, as represented by the initial seven cores. The remaining four cores in the third set supported the establishment of a nuclide fraction for the special areas involving the various trenches and areas which were confirmed (or expected) to contain TRUs. (See Table 2-8.)

The data reported in Table 2C-2 of the LTP is a combination of the 43 additional cores plus the eight cores from the third data set. The core activities were reported with no geometry correction in Attachment 8 of EC 010-01. The core activities were then geometry corrected for use in the Et calculation (Attachment. 5 of EC 010-01), and the geometry corrected data were presented in the LTP Table 2C-2 except for the activated concrete samples (Sample # 3-1A, 3-2A, and 3-3A) which were used only for activated concrete characterization.

E. Net Count Rate The net count rate data were determined by counting the cores in a low background area following their removal from the building floors. The count rate values were adjusted for ambient area background, and the net cpm was reported in Table 2C-2.

MYAPC License Termination Plan Attachment 2H Revision 3 Page 1 of 13 October, 15, 2002 ATTACHMENT 2H Forebay and Diffuser Characterization Discussion

MYAPC License Termination Plan Attachment 2H Revision 3 Page 2 of 13 October, 15, 2002 Forebay and Diffuser Characterization Discussion

1. Physical Description of the Forebay/ and Diffuser The principal forebay structure consists of the forebay basin which is approximately 400 feet in length with a granite floor, rock and soil walls (or dikes), and concrete structures at both ends.

The forebay is aligned generally in a north-south direction such that the concrete structures are located at the north and south ends with the dikes forming the east and west sides. The seal pit is at the northern end, and the diffuser intake structure is located at the southern end. During operations, plant cooling water discharged into the seal pit and then flowed over a concrete seal pit weir wall, into the forebay basin. With the cooling water system permanently secured, the flow in and out of the forebay is influenced primarily by tidal fluctuations. The forebay connects to the Back River through the diffuser piping. The intake to the diffuser piping is at the southern end of the forebay. See Figure 2H-1.

The forebay dikes were designed and constructed to achieve structural stability and minimize leakage by the choice, dimensions, and placement of pervious, impervious, and protective materials. On the interior sides of the dikes (that is, on the forebay side), the exterior layer consists of two feet (or greater) of large protective coarse rock (rip-rap). Beneath the rip-rap is about two feet of cobble stones1. Underneath the cobble stone layer is about two feet of gravel (pervious fill). Finally, beneath the gravel layer is impervious fill material. The dike walls are inclined at a slope of approximately 1.75:1 (that is, 1.75 feet horizontal run for every 1 foot of vertical drop) which results in a slope angle of about 30 degrees from the horizontal plane. See Figure 2H-2.

The diffuser system consists of large fiberglass pipes which connect the forebay basin to the diffuser discharge, submerged in the Back River. At the forebays southern end, the diffuser supply piping is nine feet in diameter. Downstream sections continually decrease to a diameter of approximately 5 feet with nozzles of 18 inches in diameter, spaced in the diffuser discharge piping. The diffuser at its discharge is submerged at a depth of over 40 feet below MSL.

The characterization of the forebay identified the following principal contaminated media:

• Floors of the forebay and seal pit. This includes other concrete surfaces, such as the seal pit weir wall. (This weir wall will be demolished down to 3' below grade.)

• Rip-rap, contaminated on the rock surfaces.

• Marine sediment (primary organic material), deposited on floors of forebay basin and seal pit and around the rip-rap.

1 This 2 foot thick layer is specified to be 6 inch minus, i.e., containing material no greater than 6 in diameter. In Figure 2H-2, this layer is referred to as fine rock cover.

MYAPC License Termination Plan Attachment 2H Revision 3 Page 3 of 13 October, 15, 2002

• Dike soil, that is, any material interior to dike below the rip-rap covering, including cobble, gravel, and other soil materials, as well as sediment deposited around the cobble.2 Remediation plans call for the removal of a majority of the accessible marine sediment in the forebay. Once the sediment remediation is accomplished, the principal contamination source term is expected to be the dike soil beneath the rip-rap, based on the assessment of activity levels in the various media. As noted above, the other contaminated media that would remain are the rip-rap (with surface contamination) and whatever sediment and other surface contamination that may remain on forebay/seal rock and concrete floors. See Section 6.6.9 for the discussion of the dose assessment and contribution of each of these remaining contaminated media.

The characterization of the diffuser identified two principal contaminated media, namely:

• Marine sediment that has been re-deposited internal to the diffuser piping by tidal action (following the permanent shutdown of the plants cooling water system).

• Contaminated internal surfaces of the diffuser fiberglass piping.

Seaweed is also considered in the diffuser dose assessment; therefore, characterization information is discussed in this attachment. See Section 6.6.9 for the dose assessment related to diffuser source terms.

2. Forebay (and Seal Pit): Contaminated Media Characterization As part of the sites initial characterization (by GTS-Duratek), several forebay samples were obtained and analyzed. Subsequent to that sampling (late 2000), an additional set of 15 sediment samples were obtained by Maine Yankee (see EC 004-01), composited, and analyzed for HTDs.

The LTP Rev. 1 nuclide fraction for forebay sediment (Section 2.5.3.e) was established based on this sampling and analysis (decay corrected to 1/1/2004). No TRUs were detected in this 2000 composite sample.3 This nuclide fraction is presented in Table 2H-1 below.

In 2001, an expanded sampling program was developed and implemented to support further characterization and remediation planning. This effort involved more extensive sampling of the forebay and principal forebay features to gain insight regarding spatial variations in activity, sediment deposition, and the activity depth profile interior to the forebay dikes. At the same time, remediation planning was involved in a number of studies and field tests to determine the optimum remediation techniques. These studies and tests also included the evaluation of material handling equipment required to address the somewhat unique challenges of the forebay, given the marine environment, variety of material sizes (from rip-rap to glacial till), and relatively steep slopes (of the forebay dikes).

2 An additional, extensive dike boring program was completed in the third quarter of 2002 to better define remediation requirements of the dike soil beneath the rip-rap. The NRC will be updated on Maine Yankees assessment of these results by the end of 2002.

3 The composite forebay sediment sample was analyzed for a standard suite of TRU nuclides. See Attachment 1 of EC-041-01 for identification of specific nuclides.

MYAPC License Termination Plan Attachment 2H Revision 3 Page 4 of 13 October, 15, 2002 Table 2H-1. Forebay Sediment Nuclide Fraction (Decay corrected to 1/1/2004)

Nuclide Fraction Co-60* 0.567 Cs-137* 0.030 Sb-125 0.005 Fe-55 0.165 Ni-63 0.233

• The resulting Co-60/Cs-137 from this data is 18.9.

The 2001 sampling program included the following principal tasks:

• Sampling of organic sediment around the rip-rap on both the east and west dikes;

• Sampling of sediment material accumulated on exposed rock surfaces in the vicinity of the weir wall at the northern end of the forebay;

• Sampling of underwater sediment on forebay basin floor and on the bottom (floor) of the seal pit.

• Subsequent, depth profile sampling into the dike material or soil.

In addition, as part of work directly related to remediation planning, rip-rap surface samples were analyzed for material composition and activity concentration.

The results of the characterization efforts are summarized below. See EC-041-01 for additional detail on sample locations, individual sample results, analysis of results, and use in the dose assessment 2.1 Dike Spatial Activity Distribution A total of forty (40) sediment samples were taken to provide information of spatial variance of activity in the sediment deposited in the tidal zone around the rip-rap on the forebay dike interior surfaces. Twenty (20) samples were obtained on each dike, i.e., ten samples along the high tide line and ten (10) samples along the low tide line. See Table 2H-2.

MYAPC License Termination Plan Attachment 2H Revision 3 Page 5 of 13 October, 15, 2002 Table 2H-2. Sediment Around Rip-Rap at Forebay High & Low Tide Lines Co-60 (pCi/g) Cs-137 (pCi/g)

Sample Location Max Min Avg Std Max Min Avg Std Co/Cs Dev Dev Ratio High Tide Line 92.6 1.8 16.9 21.8 6.5 0.2 1.2 1.4 14.6 (East & West Combined)

Low Tide Line 62.7 4.5 22.5 15.2 1.9 0.3 1.0 0.5 22.0 (East & West Combined)

High & Low Tide Line 92.6 1.8 19.7 18.8 6.5 0.2 1.1 1.0 18.0 Combined As shown in Table 2H-2, the sediment samples collected at the low tide line reported a higher Co-60 average than those collected at the high tide line. (The Cs-137 values for both high and low tide were relatively low by comparison.)

The two tidal area sediment samples with the highest reported Co-60 activity were 63.6 and 92.6 pCi/g, collected on the northern portion of the west dike at high tide. See Figure 2H-1. Because of the high concentrations, these particular locations were chosen for additional sampling to explore the activity profile interior to the dikes. The results from this effort are described below in Section 2.4 (of this attachment).

While these levels in the tidal area sediment are high relative to remediation levels (i.e.,

the DCGL proposed in Section 6 dose modeling), the profile sampling confirmed at these locations that a large portion of the contamination is near the dike soil surface, that is, the material immediately beneath the rip-rap covering. The profile sampling also showed that contamination levels, as expected, decreases rapidly with depth. Since the contaminated sediment is generally accessible, loose, and concentrated near the surface, measures under consideration for sediment remediation around the rip-rap and on the basin/seal pit floors are expected to be quite effective.

Dose modeling addresses each of the contaminated media (described in Section 1 of this attachment) including separate treatment of contaminated floors and the interior dike soil.

See Section 6.6.9.

2.2 Exposed Sediment Material (in vicinity of weir wall)

Nine (9) samples were collected from material (sediment, soil, and other material) available on the exposed rock, i.e., having no rip-rap layer, at the northern end of the forebay/seal pit structure in the area of the seal pit weir. Most of these samples were obtained on the west side to provide appropriate coverage of the area in the path of the

MYAPC License Termination Plan Attachment 2H Revision 3 Page 6 of 13 October, 15, 2002 emergency spillway.4 This set of exposed sediment samples exhibited the highest activity concentrations of all samples obtained in this particular sampling campaign of Spring 2001. See Table 2H-3 a summary of these results.

Table 2H-3. Sample Results: Sediment from Exposed Rock Surfaces and Underwater Sediment Co-60 (pCi/g) Cs-137 (pCi/g)

Sample Location Max Min Avg Std Max Min Avg Std Dev Dev Exposed Sediment 445.0 0.2 65.9 148.3 23.8 0.3 3.3 7.7 Material Underwater Sediment 62.7 5.5 19.0 16.4 7.0 0.2 1.9 2.1 (Forebay and Seal Pit)

a. The average activities of the exposed sediment material samples were: 65.9 pCi/g Co-60 and 3.3 pCi/g Cs-137. The maximum reported activity, 445 pCi/g Co-60 and 23.8 pCi/g Cs-17, was associated with a sample collected on the western side, near the weir. See Figure 2H-1 for approximate location. The second highest sample, collected from an area immediately adjacent to the above sample (on the exposed rock), reported 130 pCi/g Co-60 and 3.3 pCi/g Cs-137.

b. Not only did these samples report the maximum activity for any location sampled in this campaign, but also they were particularly high relative to the other exposed sediment samples. For sample the Co-60 concentrations ranged from 0.2 to 10.7 pCi/g and Cs-137 from 0.3 to 0.5 pCi/g for the other seven (7) exposed sediment samples.

c. The average exposed sediment sample activities (excluding the two highest samples) were 2.64 pCi/g Co-60 and 0.4 pCi/g Cs-137. The average activities for all nine (9) exposed sediment samples were 65.9 pCi/g and 3.3 pCi/g Cs-137. The average Co/Cs ratio was 19.8 (using the data from all nine samples).

d. The two highest exposed sediment samples were sent to an outside laboratory for HTD analyses. The nuclide fraction results from these HTD analyses were 4

From 1972 until late 1974, cooling water discharge passed over the weir and directly into Bailey Cove.

During that time period, the flow path included portions of exposed rock now part of the western dike (at the northern end). Construction of the west dike and diffuser system was completed in 1975. The western exposed rock then became part of an emergency spillway to provide a pathway in the event the diffuser system was not operating properly.

MYAPC License Termination Plan Attachment 2H Revision 3 Page 7 of 13 October, 15, 2002 comparable to the 2000 composite sediment HTD results with the exception that the exposed sediment sample analysis identified the presence of TRU nuclides in very low concentrations. The MDC values of the 2000 composite sediment sample analyses would have been low enough to detect the TRU nuclides had they been present at the levels found in the later exposed sediment samples.5 The original and later HTD data sets were compared and evaluated. The TRU nuclides, reported in the exposed sediment samples, were determined to represent less than 1% of the total dose associated with forebay media and were, therefore, eliminated from the nuclide fraction. Overall, it was determined that the original nuclide fraction for sediment (reported in LTP Rev. 1) was conservative due to the its higher proportion of dose-significant gamma emitters (i.e., Co-60, Cs-137, and Sb-125). The original nuclide fraction was, therefore, used in the dose assessment.

e. Lastly, as mentioned above, the exposed rock area, by its nature, contains only a small amount of material. While two of the exposed sediment samples reported very high activity, it is expected that remediation measures in this area will be quite effective because the total volume of material on these exposed rock surfaces is relatively small and because the contamination is loose and accessible.

2.3 Underwater Forebay (and Seal Pit) Sediment Thirteen (13) sediment samples were taken from underwater areas in the forebay and seal pit. Activity levels for underwater sediment were comparable to that of sediment deposited on the dikes around the rip-rap, presented in Table 2H-2 above. The overall average activities (combining forebay and seal pit samples) are 19.0 pCi/g Co-60 and 1.9 pCi/g Cs-137. Table 2H-3 summarizes the results from this sampling. Since this sediment is accessible (by diving operation) and can be vacuumed by any number of proven techniques, remediation measures for this contaminated media are expected to be quite effective.

2.4 Dike Soil Activity Profile As discussed in Section 2.1 above, depth profile samples were taken at the two locations exhibiting the highest activity levels in the rip-rap tidal zone. This sampling was undertaken to gain further insight regarding the penetration of activity into the dike interior (and to support remediation planning). See Figure 2H-1 for the surface (starting) location for these profile samples.

The depth profile samples were taken in 6 intervals down to a depth of 24." The dike 5

See Attachment 1 of EC-041-01 for the listing of MDC values obtained in the subject sediment analyses by Duke Engineering and Services Laboratory, i.e., the 2000 composite forebay sediment sample and the more recent, higher activity exposed sediment samples (Sample numbers: H059 and H060).

MYAPC License Termination Plan Attachment 2H Revision 3 Page 8 of 13 October, 15, 2002 soil material for each 6" interval was composited. Both series demonstrated a generally decreasing activity concentration with depth. See Table 2H-4, which provides the average Co-60 and Cs-137 activities values (average of the two profile samples at a given profile location). Overall, this data indicates that the majority of the contamination is concentrated near the surface of the dike soil.

It is recognized that additional sampling of the dike soil is appropriate for remediation planning and to confirm activity level assumptions used in the dose assessment. This sampling effort involved the use of boring into the area beneath the rip-rap (parallel to the slope) by way of inclined drilling from the top of the dike. This boring sampling campaign has been completed with the results currently under evaluation. This sampling effort is considered confirmatory in nature and of most value to remediation planning.

The NRC will be updated on the results of this effort.

Table 2H-4. Depth Profile Sample Results6 (Average activity values for samples collected at the listing location)

Co-60 Cs-137 Co/Cs Location pCi/G pCi/g Ratio Surface sediment7 78.2 4.6 17.0 6 (Composite)8 15.3 0.9 16.7 12 (Composite) 6.7 0.6 11.0 18 (Composite) 2.6 0.3 8.2 24 (Composite) 2.8 0.2 12.0 6

Depth profile samples were collected at the location of the highest reported activities for sediment collected beneath the rip-rap in the tidal zone, i.e., surface sediment. See Section 2.1 in this attachment.

7 Surface sediment activities, presented here for comparison, are the averages of the two sediment samples, collected immediately beneath the rip-rap, which reported the highest activity.

8 These activities represent an average of the two samples taken at the listed interval, for example, dike soil collected and composited from the 0 to 6 interval.

MYAPC License Termination Plan Attachment 2H Revision 3 Page 9 of 13 October, 15, 2002 2.5 Rip-Rap Rock, Surface Activity As part of other remediation planning activities (mentioned above), material samples were obtained from rip-rap rock surfaces. The contamination was noted to adhere to the rip-rap rock surface much like that on diffuser piping surface, i.e., by being incorporated into an organic film. The surface material adhering to the rip-rap (in areas exposed to tidal action) exhibited the same general appearance as that found on the piping coupons retrieved for analysis from the diffuser piping. The surface activity concentrations (on rip-rap and diffuser piping) were also comparable. For these reasons, the rip-rap surface data and the information from the diffuser piping surfaces were used to establish the average rip-rap rock surface activities of 0.1 pCi/g Co-60 and 0.1 pCi/g Cs-137. Table 2H-5 lists the rip-rap surface activities and offers comparison to other media contamination levels.

2.6 Forebay/Seal Pit Floors and other Forebay Concrete Surfaces No contamination data is available for the forebay/seal pit floors (or other forebay concrete surfaces). The largest surface area is represented by the forebay basin floor which consists of a granite ledge with a relatively low permeability and rock fill.

Remediation methods expected for these surfaces are expected to be highly effective.

Contamination levels for these surfaces will be confirmed as part of the remediation process. From a dose assessment standpoint, a conservative surface contamination level (DCGL) was established to bound any contamination that may remain on the forebay/seal pit floor surfaces. See LTP Section 6.6.9.

Table 2H-5. Summary Media Activity Data for the Forebay/Seal Pit (for the Principal Nuclides)

Co-60 Cs-137 Comment pCi/g pCi/g Forebay floor TBD TBD Expected to be largely remediated with remediation of (and limited marine sediment. Conservative surface contamination concrete level assumed in dose assessment.

surfaces)

Rip-rap rock 0.1 0.1 Based on both diffuser and rip-rap rock surface samples.

surface (Co-60/Cs-137 ratio: Approx. 1.0)

Marine 19.7 1.1 Marine sediment is expected to be largely remediated in sediment near the initial stage of forebay/seal pit remediation. (Co/Cs-rip-rap9 137 ratio: Approx. 18.0) 9 Average of sediment samples collected beneath rip-rap. See Table 2H-2.

MYAPC License Termination Plan Attachment 2H Revision 3 Page 10 of 13 October, 15, 2002 Table 2H-5. Summary Media Activity Data for the Forebay/Seal Pit (for the Principal Nuclides)

Co-60 Cs-137 Comment pCi/g pCi/g Dike soil 7.3 0.6 Material beneath rip-rap. May be remediated, depending material10 of results of future remediation planning. (Co-60/Cs-137 ratio: Approx. 12.1)

3. Diffuser, Contaminated Media Characterization As noted above, the principal diffuser contaminated media included: (1) marine sediment likely redeposited back into the diffuser discharge piping (following the permanent shutdown of the plant circulating water system) and (2) the diffuser piping internal surfaces. From a dose standpoint, the principal dose contributor is the marine sediment entrained in the diffuser. The plant derived activity in this sediment originated in the plants licensed liquid effluent releases (via the forebay). Then, with the securing of plant operations and the cooling water system, the tidal action transported benthic silt back into the diffuser system. Plant derived activity concentrations reported for marine sediment now inside the diffuser piping are higher than that measured in sediment outside the piping.11 The higher sediment activity inside the piping is believed to be due to activity absorbed or incorporated into the sediment inside the piping from the liquid effluent discharges since the end of plant operations. Although the dose consequences of the licensed liquid effluent releases which resulted in the activity in the diffuser have already been accounted for and reported in the routine effluent release reports, a dose assessment of the activity conservatively assumed to remain in the diffuser is discussed in Section 6.6.9.

As a matter of completeness in this discussion, seaweed characterization data is also included here since it is considered as a potential contaminated media in the dose pathway analysis. See the discussion below.

3.1 Diffuser

Marine Sediment Inside Diffuser Piping During diving operations and inspections of diffuser discharge piping, sediment samples were obtained and analyzed by gamma spectroscopy. This analysis provided the following average activities are included in Table 2H-6.

10 Dike soil contamination presented here are the highest average of the two depth profiles (i.e., average of four composite samples collected to a depth of 24.

11 Per LTP Table 2B-5, Package R2000, samples taken near the diffuser reported a maximum Co-60 activity of 0.12 pCi/g.

MYAPC License Termination Plan Attachment 2H Revision 3 Page 11 of 13 October, 15, 2002 Table 2H-6. Diffuser Related Characterization Summary12 Co-60 Cs-137 Comment pCi/g pCi/g Sediment inside 1.1 0.15 Average activity. These sediment samples diffuser discharge were also analyzed for HTDs. No HTD piping nuclides were detected. See EC 041-01.

(Co-60/Cs-137 ratio: Approx. 7.3)

Diffuser inside 0.1 0.1 Average diffuser piping coupon activity.

piping surface (Co-60/Cs-137 ratio: Approx. 1.0)

Seaweed 76.8 5.63 Average activity from forebay samples (as a conservative measure). See discussion in text. (Co-60/Cs-137 ratio: 13.6) 3.2 Diffuser Surfaces During the above mentioned diving inspections of diffuser piping, coupons of the fiberglass piping were obtained and analyzed for surface contamination. The nuclides detected were Co-60 and Cs-137 at nearly equal activity. The activity levels detected were very near the MDA of 0.1 pCi/g for each nuclide and appeared to be present on the surface as a tightly adhered, thin film of organic material. The physical appearance of this material on the piping surface was similar to that noted on the contaminated rip-rap surfaces. The activity levels of the diffuser piping surface was also comparable to that on the rip-rap, suggesting similar physical mechanisms for adhering and incorporation of contamination at work.

3.3 Seaweed Activity, Relevant to the Diffuser Dose Assessment Seaweed is present in the forebay and shoreline areas around Bailey Point. Dose contributions via contaminated seaweed were considered in the diffuser dose model as a matter of completeness, even though the dose contribution was expected (and confirmed) to be low. Seaweed samples taken from shoreline locations have shown sporadic and low activity levels of radionuclide uptake. Seaweed samples taken from the forebay were used in the dose assessment as a conservative measure of any seaweed related dose.13 See Section 6.6.9 for seaweed use, pathway assumptions, and dose results. The seaweed activity values presented in Table 2H-6 are associated with forebay samples but were 12 See Attachment 3 of EC-041-01 for additional detail regarding diffuser characterization sampling, such as number of samples and individual results.

13 Seaweed and other vegetative matter in the forebay will be removed during the sediment remediation work.

MYAPC License Termination Plan Attachment 2H Revision 3 Page 12 of 13 October, 15, 2002 applied to the diffuser dose assessment.

4. Nuclide Fraction for Forebay/Diffuser Material In summary, characterization samples were obtained and analyzed from contaminated media associated with the forebay/seal pit structures, including sediment under water and around the rip-rap, material on exposed rock (near the weir), dike soil beneath the rip-rap, and rip-rap surfaces. Additional samples were taken and analyzed from sediment inside the diffuser piping, as well as material deposited on diffuser piping internal surfaces. HTD analyses were performed on 3 collections of sediment sampling sets: an earlier (MY) composite of 15 samples, two high activity samples from the exposed sediment material, and sediment collected from inside the diffuser piping. An examination of these results concluded that the original HTD sample set, used to establish the LTP Rev. 1 nuclide fraction are appropriate and conservative nuclide fractions. The sample analyses also consistently confirmed that Co-60 and Cs-137 were the principal nuclides of interest. As noted in Table 2H-7, the Co/Cs ratios for the various contaminated media are comparable, spanning the range of 10.1 to 19.8.

The Co/Cs ratios were, in general, found to be lower for lower activity samples, as would be expected. This was seen in the assessment of contamination on rip-rap and diffuser piping surfaces, as well as deeper dike soil samples. However, the use of a nuclide fraction with a much higher Co/Cs ratio, such as that in Table 2H-1, is conservative from a dose standpoint. See EC 041-01 for additional discussion.

Table 2H-7. Comparison of Co/Cs Ratios pCi/g pCi/g Co/Cs Co-60 Cs-137 Ratio LTP Rev. 1 forebay sediment NF (Table 2H-1) NA NA 18.9 Sediment around rip-rap in tidal zone (Table 19.7 1.1 18.0 2H-2)

Exposed sediment material (Section 2H-2.2a) 65.9 3.3 19.8 Underwater sediment, forebay and seal pit 19.0 1.9 10.1 (Section 2H-2.3)

Dike Soil, underneath rip-rap, average of 15.3 0.9 16.7 highest activity samples (most shallow samples taken at 6" depth)

The forebay dose assessment confirmed that nuclides other than Co-60 and Cs-137 represent only a small fraction of the dose contribution.

MYAPC License Termination Plan Attachment 2H Revision 3 Page 13 of 13 October, 15, 2002 Thus, considering the overall dominance of Co-60 and Cs-137 nuclides in the dose impact, the comparable Co/Cs ratios for forebay/diffuser materials, and the effective absence of TRU nuclides, an overall evaluation of this characterization data concluded that a single nuclide fraction, determined by HTD analyses was appropriate for application to forebay/diffuser media.

Further assessment and comparison of the HTD analyses concluded that the originally determined nuclide fraction, established in the LTP Rev. 1 analysis of forebay sediment, remained appropriate and conservative for dose assessment application to forebay and diffuser contaminated media. See EC 041-01 for additional detail and discussion of the data evaluation.

MYAPC License Termination Plan Attachment 2I Revision 3 Page 1 of 5 October 15, 2002 ATTACHMENT 2I Soil Sampling and Radionuclide Fraction

MYAPC License Termination Plan Attachment 2I Revision 3 Page 2 of 5 October 15, 2002 Soil Sampling and Radionuclide Fraction1 Introduction Multiple soil samples representing areas of the site known to have high activity soil contamination were collected. Several samples from each area were composited to provide the most representative contaminated soil values and provide the highest probability to detect and quantify hard-to-detect (HTD) radionuclides that could be associated with the contaminated soil.

Specific instructions were included for composition and analysis of these samples so as to insure the representation of the samples to be submitted for HTD vendor lab analysis. Since the final status surveys for soil include gamma spectroscopy analysis of each soil sample, the HTD data set is useful for establishing the surrogate relationship to Cs-137. These HTD nuclides (H3 and Ni-63) contribute to less than one percent of the total soil dose.

Sample Analysis A comparison of specific soil nuclide parameters over time (1999- 2030) was made to determine how the soil Cs-137 surrogate DCGL value changes with time. The DCGL ranges from about 4.2 to 4.4 pCi/g. The change is mostly due to the fact that Co-60 decays at a faster rate than Cs-137, which results in higher allowable surrogate DCGL levels at later times. This variation with respect to time shows that the effect of conducting final status surveys significantly sooner or later than the currently proposed time is insignificant. In practice, the Co-60 will be measured by gamma spectroscopy and the only nuclides included in the Cs-137 surrogate calculations will be H-3 and Ni-63. The total dose from H-3 and Ni-63 in soil is about 0.1 mrem/y as calculated for the year 2004. Any changes in dose from these radionuclides over time will be negligible (<0.02 mrem/y) relative to the unrestricted use criteria.

Sample Selection and Composition To determine the best representation of Industrial and Restricted Area samples the soil samples and respective locations collected during the GTS Site Characterization were examined.

Emphasis was placed on samples collected from areas of principle spill or contamination incident. These areas of significance were the RWST, PWST and the Shielded Radiological Waste Storage Area (SRWSA). Examination of all other site characterization soil samples showed that these three areas contained the maximum concentration of elevated soil activity.

The available GTS vendor laboratory results for some samples from these areas showed relatively high MDA values for several HTD nuclides. Any positive TRU results were at or very near MDA values and those near the MDA value did not appear in the ratios of one nuclide to another, which would be expected in power reactor TRU inventory. From these observations it was decided to composite biased samples of maximum concentration from the regions of the most significant incidents. The twelve samples that were composited for these areas originated 1

The soil sample analysis results and general methodology are presented in Engineering Calculation EC 013-01, Rev. 0. This calculation reviews the associated sample results and encompasses the features and nuclides associated with Engineering Calculation EC 007-00, Rev 1.

MYAPC License Termination Plan Attachment 2I Revision 3 Page 3 of 5 October 15, 2002 from the archived GTS site characterization soil samples and are expected to represent greater location diversity and better estimate of distribution than individual sample locations.

Inspection and composite instructions were developed in the form of a technical evaluation document so that all samples were systematically processed in the same manner and using the same methodology. Once the twelve archived samples were located, the samples were assigned a new chain-of-custody form and a minimum of one sample from each group (RWST, PWST and SRWSA) was analyzed in the original GTS retrant beakers using the Maine Yankee gamma spectroscopy system. As one of the instruction steps, the Maine Yankee spectroscopy Cs-137 analysis results were compared to the original GTS Cs-137 results and found to reasonably agree. As stated, the Maine Yankee analysis results conclude that the principle gamma emitters associated with the original GTS soil sample containers were within reasonable agreement of the concentrations reported in the GTS Characterization Report.

Following this comparison and per the composite instructions each of the GTS samples for each of the three regions (RWST, PWST and SRWSA) were thoroughly mixed and a predetermined sample mass collected of the composite representing each region. From the original Cs-137 concentration associated with each sample, the concentration per unit mass and total mass of the sample was estimated. These results were compared to the composited sample results. This comparison provides both a final check of the reported concentrations to the current analysis and insight into the distribution of associated radionuclides in the media. The narrow range of concentration variation associated with the RWST estimated and final composite values is indicative of contaminants associated with liquids where the concentration would expectedly be more uniform. The wider range of variation for the PWST and Shielded Storage areas estimated and final composite values are indicative of non-uniform contaminants and for a given sample group the range variation would represent the spatial distribution of the activity in the media.

Table 2I-1 presents these findings.

Table 2I-1 Original and Composite Cs-137 Soil Concentrations and Comparison GTS Samples Estimated Collective Final Composite Value Value Sample Location Cs-137 Range Weight Cs-137 Weight Cs-137 (pCi/g)

(g) (pCi/g) (g)

RWST 11.0 -114.0 1440.0 61.6 1475.0 60.5 PWST 14.6 - 156.0 1500.0 86.1 1532.0 99.4 Shielded Storage 18.3 800.0 18.3 1023.0 22.1 All RWST samples represent surface soil; Two PWST soil samples (Cs-137 ranging from 14.6 - 57.6 pCi/g) represent soil at 6-18 inch depths. Three PWST sample represent Cs-137 surface soil ranging from 69.1-156 pCi/g. The single Shielded Storage sample is surface soil.

MYAPC License Termination Plan Attachment 2I Revision 3 Page 4 of 5 October 15, 2002 Additional Confirmatory Sample Collection The continued characterization soil samples (442 samples) collected in 1999 and 2001 support the observed composite results obtained from the samples associated with the RWST, PWST and SRWSA.

These investigations focused primarily in the Industrial and Restricted Area of the site. Only Co-60 and Cs-137 were identified in the 442 samples. These samples represented both surface and subsurface investigations to a depth of nearly 4 meters (~12 feet). The concentration range of all these samples was significantly lower than the samples used for the soil profile provided in EC-013-01 (See Table 2I-2 below).

A total of 442 samples from 107 locations were collected and analyzed. The sample analysis results (442 samples) showed that Cs-137 was reported at >MDA 35.5 percent of the time while Co-60 was reported at >MDA only 2.0 percent of the time. The results of these samples provide additional support for Cs-137's predominate presence in contaminated site soils. DCGL values show that the surrogate DCGL changes little over time (~2.6% from 2004 to 2030). The maximum observed soil concentrations for Cs-137 and Co-60 in the 1999 and 2001 sample results (442 samples) were considerable lower (Cs-137: 34.7 and Co-60 12.4 pCi/g) than the composited samples used to determine the HTD soil constituents. These results indicate that the analyzed composites conservatively address the HTD and gamma emitters associated with the site soils. Of the 422 samples 79 were determined to exceed the action level estimated for the sampling plan.

Table 2I-2 presents the range of Co-60 and Cs-137 for the 79 samples that were found to exceed the sample plan respective Action Levels of 1.0 and 3.1 pCi/g for Co-60 and Cs-137. It is important to note that for the Co-60 data in Table 2I-2 only seven Co-60 sample results are above the MDC for the analysis parameters (The reported Co-60 MDCs for the remaining 72 samples ranged from 0.05 to 0.40 pCi/g). For the Cs-137 data in Table 2I-2 a total of 58 (73.4%)

of the 79 samples are above the MDC for the analysis parameters (The 21 Cs-137 samples less than the MDC ranged from 0.06 to 0.41 pCi/g). The results of Table 2I-2 show that none of the 442 additional samples collected approached the soil concentrations reported for the RWST and PWST composite samples. As previously stated the radionuclide results of the RWST and PWST samples conservatively characterize the Maine Yankee site soils.

MYAPC License Termination Plan Attachment 2I Revision 3 Page 5 of 5 October 15, 2002 Table 2I-2 Cs-137 and Co-60 Range for Continued Characterization $Action Level*

Cs-137 Number Co-60 Number Range (pCi/g) of Observations Range (pCi/g) of Observations

>0.34 -2.0 33 >0.06 - 0.50 36

>2.0 - 5.0 22 >0.50 - 1.0 27

>5.0 - 10.0 15 >1.0 - 2.0 15

>10.0 - 20 3 12.4 1

>20 - 34.7 6 Total 79 Total 79

• Sample Plan Action Level 1.0 and 3.1 pCi/g for Co-60 and Cs-137 respectively.

Summary

• The soil characterization by GTS and sample locations throughout the Restricted Area (RA) and Industrial Area (IA) were reviewed. Sample locations were selected that reflected locations of historic primary contamination incidents and highest soil contamination.

• The concentrations of the selected samples increase the probability of detecting and quantifying HTD nuclides.

• The composite method used resulted in composite soil concentrations conservatively higher than any of the GTS characterizations soil samples and the 442 continued characterization samples acquired in the RA and IA in1999 and 2001.

• All FSS soil samples will be analyzed using gamma spectroscopy.

• The 442 continued characterization soil samples collected in 1999 and 2001 support the composite results.

• The Cs-137 surrogate DCGL for soil varies no more than 2.6% from 2004 through 2030

(~4.0% from 1999 through 2030).