ML022810429

Final Report RAE-42613-003-4R1, Calculation of Sub-Surface Dcgls for the Saxton Nuclear Experimental Corporation Site

ML022810429
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Site:	Saxton File:GPU Nuclear icon.png
Issue date:	09/20/2002
From:	URS Corp
To:	Office of Nuclear Reactor Regulation
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RAE-42613-003-4R1
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RAE-42613-003-4R1 FINAL CALCULATION OF SUB-SURFACE DCGLs FOR THE SAXTON NUCLEAR EXPERIMENTAL CORPORATION SITE September 20, 2002 URS CORPORATION 756 East Winchester Street, Suite 400 Salt Lake City, UT 8410

RAE-42613-003-4R1 FINAL CALCULATION OF SUB-SURFACE DCGLs FOR THE SAXTON NUCLEAR EXPERIMENTAL CORPORATION SITE Prepared for GPU Nuclear, Inc.

Route 441 South Middletown, PA 17057-0480 September 20, 2002 URS CORPORATION 756 East Winchester Street, Suite 400 Salt Lake City, UT 84107

TABLE OF CONTENTS Chapter Page No.

I INTRODUCTION 1-1 1.1 Contamination History 1-1 1.2 Objective and Scope 1-4 2

CONCEPTUAL MODEL 2-1 2.1 Site Hydrogeology 2-1 2.2 Areas of Concern 2-4 2.2.1 Steam Plant Area 2-4 2.2.2 General SNEC Area 2-6 2.3 Radiation Exposure Scenarios and Transport Pathways 2-7 3

CALCULATION MODELS 3-1 3.1 Steam Plant Backfill 3-3 3.2 General Site Overburden 3-5 3.3 Bedrock 3-7 3.4 Methodology 3-7 4

ANALYSIS METHODS AND RESULTS 4-1 4.1 Steam Plant Backfill 4-1 4.2 General Site Overburden 4-3 4.3 Bedrock 4-5 4.4 Kd Uncertainty 4-7 5

SUMMARY

AND CONCLUSIONS 5-1 REFERENCES R-1 APPENDIX A RESRAD Input Distributions A-1 ii

LIST OF FIGURES Figure Page No.

1-1 SNEC site layout 1-2 2-1 Conceptual representation of the hydrogeology at the SNEC site 2-2 2-2 Areas of concern at the SNEC site 2-5 2-3 Radiation exposure pathways 2-8 2-4 Basement excavation model for general site area 2-11 2-5 Basement excavation model for steam plant backfill 2-12 3-1 RESRAD's water transport model 3-2 3-2 RESRAD representation of the Steam Plant Backfill 3-4 3-3 RESRAD representation of the Overburden Layer 3-6 3-4 RESRAD representation of the Bedrock Layer 3-8 3-5 DCGL Methodology 3-9 iii

LIST OF TABLES Table Page No.

1-1 Radionuclides observed in site samples 1-5 4-1 Steam Plant Backfill DCGLS (pCi/g) 4-2 4-2 General Site Area Overburden DCGLs (pCi/g) 4-4 4-3 General Site Area Bedrock DCGLs (pCi/g) 4-6 5-1 Site-Wide Composite Subsurface DCGLs (pCilg) 5-3 A-I Steam Plant Backfill Input Parameter Distributions A-3 A-2 General Area Overburden Input Parameter Distributions A-5 A-3 Bedrock Input Parameter Distributions A-7 A-4 Dose and DCGL Summary A-9 iv

1.

INTRODUCTION The Saxton Nuclear Experimental Corporation (SNEC) operated a 23.5-megawatt thermal pressurized water research and training reactor from 1962 to 1972 at its facility near Saxton, PA. As is shown in Figure 1-1, the reactor was located adjacent to a steam-turbine electric generating station that operated from 1924 to 1972. Since the reactor shutdown, GPU Nuclear, Inc. (GPU) has assisted SNEC in removing and disposing of the reactor fuel and internal parts and in characterizing and decontaminating large portions of the site.

In preparing to terminate the Nuclear Regulatory Commission (NRC) license for the site, GPU determined Derived Concentration Guideline Levels (DCGLs) for the top meter of site soil that correspond to the 25 mrem/year total radiation dose limit prescribed by NRC for site cleanup and the 4 mrem/year dose limit for drinking water. GPU contracted with URS Corporation (URS) to develop and apply a conceptual model and methodology to determine DCGLs for the sub-surface zone below the top meter of soil at the site.

1.1 CONTAMINANT HISTORY Radioactive materials are considered to have been on the SNEC Site from the time the site was licensed in 1962 by NRC to possess such materials and fuel the reactor. The operational history of the site indicates that the area (approximately 60 m x 75 m) around the reactor Containment Vessel (CV) included a Control and Auxiliary Building, a Radioactive Waste Disposal Facility, an underground pipe tunnel, a drum storage bunker, and a refueling water storage tank (GPU, 2000a,b). These facilities were decontaminated in 1987 - 1989 and all but 1-1

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the CV were demolished in 1992 after acceptance of a final release survey by NRC (NRC, 1992).

The soils removed around the CV and structures were replaced with clean backfill soil. Although there is no evidence of leakage from the CV itself, the contamination removed from the areas surrounding the outside of the CV suggests the occurrence of surface spills and leaks from buried piping and tunnels.

The Steam Plant features underground concrete intake and discharge tunnels to cycle cooling water from the nearby Raystown Branch of the Juniata River (River). Since the reactor steam contained low levels of radioactivity, further contamination occurred when the steam was utilized in the generating station. When reactor steam was utilized in the station, low levels of radioactivity were cycled through the plant discharge tunnel.

There is also a possibility that radioactive contamination occurred in the intake tunnel from warm discharge water that was recycled through the intake tunnel to avoid ice buildups during cold winter periods. Although radioactivity levels in the discharge tunnel were low enough to satisfy the radiation regulations then in effect, some radioactivity tended to accumulate in some Steam Plant structures, tunnel sediments, and surrounding soils near concrete cracks and joint leaks.

A Spray Pond was operated approximately 260 m southwest of the CV to cool water from the steam plant during summer months before release to the River. The pond consisted of arrays of pipes and spray nozzles covering a 40 m x 90 m area of surface soil. The radioactivity in the heated water could have been released and accumulated in the soils in the spray pond vicinity. Some building rubble from demolition of the Steam Plant also was disposed in the former Spray Pond area.

The primary radionuclides identified in analyses of steam plant sediments and soils in the CV vicinity are H-3, Sr-90, Co-60, Cs-137, and Am-241.

Additional radionuclides have also been observed in one or more site samples or have been hypothesized to occur in contaminated 1-3

materials, as listed in Table 1-1. All of these radionuclides will be analyzed for estimating site subsurface DCGLs.

1.2 OBJECTIVE AND SCOPE This report presents the DCGLs developed by URS for the SNEC Site materials deeper than one meter. Included in this report is a presentation of the methodology used in the analysis and a summary of the distribution of input parameters chosen to represent the SNEC Site hydrology and meteorology. URS developed these analysis input distributions from reviews of historical and current technical reports furnished by GPU and in consultation with GPU personnel and their hydrologic consultant. The DCGLs are designed to satisfy the 25 mremfyear total dose limit and 4 mrem/year drinking water dose limit for members of the general public that could receive the maximum radiation exposures from the SNEC Site and its environments.

1-4

Table 1-1 Radionuclides observed in site samples.

H-3 Ni-63 Eu-I 52 Pu-241 C-14 Sr-90 Pu-238 Am-241 Co-60 Cs-137 Pu-239 1-5

2.

CONCEPTUAL MODEL The conceptual models of the Site are based upon available site characteristics. These characteristics have been observed through hydrologic well logging activities, and in-situ and laboratory analyses of site soils. These characteristics are used to identify two representative areas of concern. Conceptual models for these two areas are summarized below.

2.1 SITE HYDROGEOLOGY Well logs show the near-surface hydrogeology to be a consistent pattern of three distinct layers of materials, (A) Fill, (B) Overburden, and (C) Bedrock (illustrated in Figure 2-1).

Previous geotechnical and hydrologic investigations provided to URS by GPU identify the characteristics of these materials (H&A, 2001).

The Fill layer near the CV, Steam Plant, and Spray Ponds has been observed to be 0.46 to 1.22 meters thick. It is represented for modeling purposes to be about 1 meter thick with a range from about 0.4 to 2 meters over the larger site area. The Fill generally consists of well-graded silty and clayey fine to coarse sand with fine to medium gravel. In some areas, it also contains a well-graded mixture of ash and cinders from the former Steam Plant. In remediated areas near the CV and Steam Plant, clean backfill from an off-site source comprises the top meter. The Fill is estimated to have a total porosity of 0.46 (range from 0.35 to 0.56), an effective porosity of 0.41 (range from 0.28 to 0.54), a field capacity of 0.136 (range from 0.079 to 0.192), and a hydraulic conductivity of 32.3 meters/year. Although generally unsaturated, higher water levels during significant rainfall events cause periods of transient saturation.

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I II SATURATED III (but practically impermeable) below G S SATURATED III Figure 2. Conceptual representation of the hydrogeology at the SNEC site.

The Overburden or boulder layer thickness is observed to range from less than 1 meter to about 3 meters and is represented for modeling purposes by about 2 meters. The Overburden features rounded boulders interspersed with a dense mixture of sand, silt and clay. The boulders consist of hard quartzite with negligible porosity. The Overburden behaves like glacial till, with a low permeability on the order of l07 cm/s (0.032 m/y). Its bulk or total porosity is estimated at 0.10 to 0.15. Hydraulic gradients in the Overburden range from 0.02 to 0.03 based on gradients between the Site, tunnel, and river. The Overburden acts as a hydraulic barrier to flow between the Fill and Bedrock in undisturbed areas.

The Bedrock consists of fractured and weathered siltstone that begins at depths of 2.1 to 5.5 meters below the surface and is believed to extend to depths of more than several dozen meters. Saturated groundwater flow in the Bedrock generally occurs along bedding planes and within its fractures. The total porosity of the Bedrock ranges from 0.21 to 0.41 and the effective porosity is measured at approximately 0.0275. Hydraulic gradients in the Bedrock range from 0.013 to 0.03 based on gradients between the Site, tunnel, and river. The hydraulic conductivity of the Bedrock for fracture flow is estimated to be 67.9 meters/year.

The site hydrology is dominated by westward flow toward the River. The Bedrock features saturated water flow in the B/C interface and in Bedrock fractures and bedding planes.

Although the Bedrock water flow probably extends throughout the deeper parts of the Bedrock, the regional gradient promotes relatively horizontal flow in the top 20 meters that drains to the River. The Bedrock is intersected by the CV, which extends approximately 15 m beneath the surface, and by the basement of the Steam Plant, which extends approximately 7.5 m beneath the surface. Utility tunnels near the CV also extend into the Bedrock, as do the plugged intake and drainage tunnels that connect the Steam Plant basement to the River. The disturbances of the Overburden layer in constructing the CV, Steam Plant, and concrete tunnels cause high-2-3

permeability zones at their interfaces with the Overburden that hydraulically connect the Fill and Bedrock.

2.2 AREAS OF CONCERN Two areas of concern were selected to represent the parts of the SNEC Site that have the greatest potential to cause present or future radiation exposure to members of the public. These areas are illustrated in Figure 2-2. They include the Steam Plant area and the remainder of the general SNEC Site (excluding the Steam Plant area). These areas were chosen to represent the site because of their potential for elevated radionuclide concentrations or their association with radiation exposure pathways.

2.2.1 Steam Plant Area The Steam Plant Area is defined to include the existing basement of the Steam Plant and the underground intake and discharge tunnels that connect the former plant basement to the River.

This area was selected because it received reactor steam, it was a conduit for discharging reactor secondary cooling water, it is hydraulically connected with the CV Area, it contains sediments and sumps with trace Cs-137 and other contaminants, and it contains channels that hydraulically connect the Fill-layer and the Bedrock.

The Steam Plant basement has been filled with demolition rubble from the former Steam Plant building and covered with 1 meter of clean fill.

Potential residual contamination of soils, debris, and ground water may remain in this area or may occur from seepage through plugged tunnels from the CV Area.

2-4

Figure 2-2. Areas of concern at the SNEC site.

2-5

The intake and discharge tunnels that connect the Steam Plant to the River affect migration of water and potential contaminants from other parts of the site. The tunnels generally feature a permeable zone along their exterior boundaries owing to less-compact backfill and removal of clays from the Overburden cobbles during tunnel construction. The intake tunnel (1.8 m x 2.4 m) is sufficiently large to intersect the Fill and Overburden layers in some areas and the Overburden Bedrock interface in others. The intake tunnel hydraulically connects the Fill-layer and Bedrock over its entire length. The discharge tunnel has similar size and acts as a permeable path that intercepts Fill-layer water from the CV Area and diverts it to the River via the permeable zone along its exterior. The discharge tunnel hydraulically connects the Fill and Bedrock layers along approximately half its length near the Steam Plant and remains in the Fill and Overburden layers in areas closer to the River. The tunnels can therefore conduct contaminants from the permeable basement of the Steam Plant into the Fill layer, the Bedrock, or both.

2.2.2 General SNEC Area The general SNEC area of concern includes the CV and Spray Pond areas. The areas of and near the CV include the CV and the former auxiliary and waste management operations.

This area is of concern because it was the original source of most man-made radioactivity at the site and because it contains backfill that hydraulically connects the Fill layer and the Bedrock.

Although the reactor core, most internal structures and auxiliary facilities, and surrounding contaminated soils have been removed and disposed, potential residual contamination of soils and ground water that originated in this area may still remain. It is therefore considered the area with greatest potential for elevated levels of residual radioactivity.

The potential hydraulic connections along CV and tunnel interfaces with native materials could allow migration of any contaminants into the Fill layer, the Bedrock, or both.

2-6

The area near the Spray Ponds include the approximate 40 m x 90 m footprint of the former Spray Pond. It is included in the general SNEC area of concern because it seasonally received cooling water from the Steam Plant and it was later covered with building rubble from the Steam Plant. The original surface soils remain in the Spray Pond area beneath building nibble.

However, the steel pipe connecting the Spray Pond to the Steam Plant has been excavated, surveyed, and removed. While the Spray Pond was built on the surface of the Fill soil layer and was originally hydraulically isolated from the Bedrock by the overburden layer, recent and planned decontamination activities may create hydrologic transport channels through the Overburden allowing any contamination to seep into the saturated Bedrock layer.

Subsurface materials for the Spray Pond and Reactor areas (excluding the CV excavations) are very similar, consisting of approximately two meters of overburden and a greater thickness of underlying bedrock.

The subsurface material in the SSGS consists of crushed, homogenized site debris that is covered with one meter of clean fill.

Because of these differences, DCGLs will be estimated for only one material (homogenized debris) in the SSGS and for two materials (overburden and bedrock) in the Spray Pond and Reactor areas.

2.3 RADIATION EXPOSURE SCENARIOS AND TRANSPORT PATHWAYS Exposures to members of the critical population group are postulated to occur to a hypothetical individual (the Receptor) who is subject to all potential exposure pathways. For both areas of concern, his exposures are considered to originate from similar exposure scenarios L

because both result from radionuclides buried in sub-surface soils (Overburden, and/or Bedrock) in locations where he could conceivably build a house and reside. The pathways for radiation exposure are illustrated in Figure 2-3.

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Atmospheric Transport URS - 107193 Figure 2-3. Radiation exposure pathways.

2-8

The Receptor is considered to reside in a home in or near areas of concern.

The most conservative parameters and their respective distributions are selected from each of the areas of concern to identify a site-wide residential scenario which results in the highest exposure. This site-wide exposure is then used to determine nuclide-specific DCGLs for each subsurface layer.

In the scenario, the Receptor is exposed to residual radioactivity in several ways that include (a) excavating and spreading contaminated Overburden material during home construction and yard leveling; (b) consuming drinking water from a Bedrock well; (c) consuming fruits and vegetables grown onsite with irrigation water from the transient flow within the Fill-soil layer; and (d) consuming beef and milk from cattle raised onsite using the same irrigation water. The shallow water table and the boulders in the Overburden layer discourage construction of a basement for the on-site residence. However, excavation and spreading of Fill material from beneath the top meter and into the upper Overburden layer could occur in leveling sloped areas for a home site.

The potential radiation exposure pathways associated with the on-site residential scenarios are analyzed to estimate radiation doses. Gamma radiation exposures occur in the yard and through the house floor from radionuclides mixed into surface soils from excavation (pathway a) and well cuttings (pathway b). Exposures from inhaling contaminated dust occur during site grading (pathway a) and well excavation (pathway b) as well as from garden tillage and wind resuspension of contaminated soils (pathways a, b, c, and d). Exposures from ingesting contaminated soil occur from soil entrained on vegetables (pathway c) and unwashed hands (pathways a, b, c, and d). Exposures from ingesting contaminated drinking water occur from transport in the Bedrock (pathway b). Exposures from ingesting contaminated fruits and vegetables occur via their uptake from contaminated surface soil (pathways a and b) and contaminated irrigation water (pathway c). Exposures from ingesting contaminated beef and milk occur from cattle fed with contaminated crops and water (pathways a, b, and d).

2-9

The basement of the hypothetical house built by the resident in the Spray Pond, Steam Plant, or Reactor areas penetrates the 1-m surface fill layer and the top part of the underlying material. The materials excavated for the basement are represented here in the same way as they have been in previous NRC guidance documents. The excavation is considered to penetrate sufficiently deep for construction of the basement, and the excavated material is considered to be mixed with overlying fill material and placed in the vicinity of the house (NRC, 1986, 1999).

Figures 2-4 and 2-5 illustrate the basement excavations and surface placement of the excavated materials for the different Site areas.

The footprint area of the hypothetical house and its basement excavation is chosen to be 200 M2, which corresponds to the house areas used in the previous NRC analyses (NRC, 1986, 1999).

The height of the basement, hb, is chosen to be 2 meters based on both the previous NRC guidance and conservative assessment of the site parameters.

The previous NRC guidance documents used basement depths of 3 meters for disposal cells that were covered by 2 m of clean soil or cap material (NRC, 1986). This resulted in 1-meter intrusion depths into the contaminated material. For the present Site, where fill soil layers are I m thick, a 2-m basement excavation provides the same 1-rn intrusion depth into potentially-contaminated material.

The Spray Pond area, Reactor area (except the back-filled CV excavation), and undisturbed areas of the Site feature a 2-in layer of boulder-laden overburden between the 1-m fill soil layer and bedrock, as shown in Figure 2-4. The overburden discourages excavation because of its large boulders and interstitial cobbles. The hypothetical 2-m basement excavation (1I-m into the boulder layer) is therefore conservatively deeper than would normally be expected.

Furthermore, the water table at the Site varies between 0.7 and 2.3 m, making the hypothetical 2 m basement excavation conservatively deeper than the average estimated water table depth.

2-10

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2-12

The Steam Plant basement contains homogenized rubble instead of overburden in the bottom half of the hypothetical 2-m excavation (see Figure 2-5). The hypothetical Steam Plant excavation therefore intrudes to the same extent (1 m) into potentially-contaminated material as in the Spray Pond area and is subject to similar water-table limitations as in other parts of the Site.

The native materials in the CV area are replaced with clean backfill to depths well beyond any realistic basement excavation depths. Therefore, any excavations in the CV area would only bring clean backfill to the surface. The surrounding materials in the Reactor area are represented by the same profile as the Spray Ponds, as illustrated in Figure 2-4.

The concentrations of any contaminants in the overburden layer beneath the Spray Ponds or in the homogenized rubble in the Steam Plant are reduced to one-half of their in-situ concentrations by mixing with the top meter of fill soil during their excavation. The two-fold dilution factor results from mixing equal volumes of clean and potentially-contaminated excavation materials using the same calculation approach advocated in the NRC guidance documents (NRC, 1986, 1999).

The mixed material from the basement excavation is considered to be spread around the hypothetical house as in the NRC models (NRC, 1986).

The area of this material is conservatively taken to be 2,000 M2 based on the nominal one-half acre (2,023 m2) and 1,000 m2 to 2,000 m 2 area ranges estimated by NRC (NRC, 1986). The average thickness of this layer of mixed material is h, = 0.20 m [estimated as (200 m2 basement footprint area) x (2 m depth) /

(2,000 m2 spread area)].

This thickness is conservative in giving an approximate maximum gamma radiation activity and approximating the default crop root depth of 0.15 m in the RESRAD dose assessment code (Yu, 2002). Additionally, the geophysical characteristics of this layer of rmixed material are depth-weighted averages of those of the Fill and Overburden layers.

2-13

3.

CALCULATION MODELS The RESRAD dose assessment model for sites contaminated with RESidual RADioactive materials is the main tool used to determine the subsurface DCGLs for the Site Bedrock and Site Overburden soils. RESRAD 6.1 was developed and adapted from earlier versions for use with the NRC Standard Review Plan (NRC, 2000) for decommissioning and as a tool for demonstrating compliance with the license termination rule in a risk-informed manner. Version 6.1 also computes probabilistic estimates of radiation dose distributions that result from various distributions of input parameters. RESRAD is used to evaluate the sensitivity of input parameters and identify parameters whose most-probable values and distributions require site-specific measurements or detailed justification. It is also used to compute radiation dose distributions from unit concentrations of radionuclides.

As is illustrated in Figure 3-1, RESRAD's basic transport model assumes three main transport zones: contaminated, unsaturated, and saturated.

RESRAD's model assumes water infiltrates into the contaminated zone and leaches radionuclides out of the waste, transporting the contaminated groundwater vertically down through the unsaturated zone and then horizontally through the saturated zone to a well.

RESRAD's representation of horizontal flow within the saturated zone assumes Darcian flow through a homogeneous, saturated, porous medium. Because the SNEC Site is highly heterogeneous, including the transient Fill and Bedrock water pathways, two basic RESRAD data sets are used in the analysis to address these layers separately.

3-1

Precipitation URS - 107196 Figure 3-1. RESRAD's water transport model.

3-2

RESRAD calculates radiation doses for a chronically exposed individual, focusing on radioactive contaminants in soil and their transport in air, water, and biological media to a single receptor. It considers nine exposure pathways: direct gamma exposure, inhalation of particulates and radon, and ingestion of plant foods, meat, milk, aquatic foods, water, and soil. Radiation doses, health risks, soil guidelines, and media concentrations are calculated for specified time intervals. The source is adjusted over time to account for radioactive decay and in-growth, leaching, erosion, and mixing.

3.1 STEAM PLANT BACKFILL RESRAD assesses exposures from the Steam Plant Backfill by assuming that contamination is brought to the surface from well drill cuttings and excavation into the Steam Plant Backfill for a house foundation and yard leveling. Exposures from ingesting contaminated meat, milk, and vegetation; inhaling dust, and direct gamma radiation exposure are evaluated.

The application of RESRAD for exposures from the Steam Plant Backfill is illustrated in Figure 3-2. Consistent with RESRAD terminology, the waste is assumed to be brought to the surface, spread, and mixed as a result of house construction, site grading, and well excavation will represent the waste zone. The region below the surface mixing zone, but above the water table is represented as the unsaturated zone (vertical transport region).

Analysis of exposures, resulting from contaminants entrained in the undisturbed Backfill are also evaluated.

Contaminants not brought to the surface as part of excavation and site leveling activities are assumed to become leached into vertically traveling groundwater. It is then assumed that these contaminants travel to a well site, where they are pumped to the surface for human and livestock drinking water, as well as crop irrigation. The peak dose and year of occurrence is then computed.

3-3

URS - 107197 Figure 3-2. RESRAD representation of the Steam Plant Backfill.

3-4

P ý The input parameters and their sources for assessment of exposures related to the Steam Plant Backfill are summarized in Table A-I of Appendix A. Site-specific values are used, whenever available. When not available, RESRAD default values are used in the analysis. Initial source concentrations of 0.5 pCi/g represent surface concentrations resulting from house construction and site excavation.

3.2 GENERAL SITE OVERBURDEN RESRAD assesses exposures from the Overburden layer at the Site by assuming that contamination is brought to the surface from well drill cuttings and excavation into the Overburden layer for a house foundation and yard leveling.

Exposures from ingesting contaminated meat, milk, and vegetation; inhaling dust, and direct gamma radiation exposure are evaluated. The application of RESRAD for exposures from the Overburden layer is illustrated in Figure 3-3. Consistent with RESRAD terminology, the waste is assumed to be brought to the surface, spread, and mixed as a result of house construction, site grading, and well excavation will represent the waste zone. The region below the surface mixing zone, but above the water table is represented as the unsaturated zone (vertical transport region).

Analysis of exposures, resulting from contaminants entrained in the undisturbed Overburden layer, are also evaluated.

Contaminants not brought to the surface as part of excavation and site leveling activities are assumed to become leached into vertically traveling groundwater. It is then assumed that these contaminants travel to a well site, where they are pumped to the surface for human and livestock drinking water, as well as crop irrigation. The peak dose and year of occurrence is then computed.

II' 3-5

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URS - 107197 Figure 3-3. RESRAD representation of the Overburden Layer.

3-6

The input parameters and their sources for assessment of exposures related to the Overburden layer are summarized in Table A-2 of Appendix A. Site-specific values are used, whenever available. When not available, RESRAD default values are used in the analysis. Initial source concentrations of 0.5 pCi/g represent surface concentrations resulting from house construction and site excavation.

3.3 BEDROCK The third RESRAD application evaluates ingestion of drinking water from a well drilled into the Bedrock. Potential waste residing near the base of the CV, Steam Plant, or transported downward from the Spray Pond Fill is modeled by RESRAD as being directly above the water saturated bedrock. As is illustrated in Figure 3-4, this zone of contamination is represented within RESRAD by assigning the vertical-transport vadose zone a negligible thickness and rapid transport properties (making the contamination immediately available to the groundwater). The input parameters and their sources for assessment of exposures related to the Bedrock layer are summarized in Table A-3 of Appendix A. Site-specific values are used, whenever available.

When not available, RESRAD default values are used in the analysis.

3.4 METHODOLOGY The technical approach and detailed steps for determining the DCGLs for the sub-surface materials at the SNEC Site are identified in Figure 3-5 based on discussions between URS and GPU personnel. The large gray numbers in Figure 3-5 correspond to the step numbers listed below.

3-7

Inhalation Inhalation Ingestion LL External Crops Wind Blown Exposure Grown on L._

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3-8 URS-107198b

' Create RESRAD Input file with distribution from H&A, D&D and RESRAD defaults Create correlations between Input 2 1 parameters Edit RESRAD file to Include 75%

value as base-determlnistic value Input for each sensitive parameter 11b that Is positively correlated (sens. parameter >0.25)

I Replace base-determinIstic value for sensitive parameters with statistical mean value for those parameters for which mean Is not 11d bounded by sens. parameters of 0 25 & 0.75 Execute RESRAD In 3Probabilistic Mode 5eExamie MC 5

sumamary report Is K0 coefficient >0.25 Is 6

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=nu ieto Include 25% value as as-eermlnlstlc lie Input for each sensitive parameter 11C Pthat In negatively correlated (sens parameter <0.25)

URS - 107262b Figure 3-5. DCGL Methodology.

3-9

1)

Generate an appropriate RESRAD 6.1 input file containing all Haley and Aldrich (H &

A) values and parameter distributions. Where available, DandD default values are used for metabolic and behavior inputs.

For parameters for which input guidance is unavailable from either DandD or H & A, default RESRAD values and distributions are used. Uncertainty analysis is performed by RESRAD for each parameter for which a distribution has been input. A list of these values and distributions is presented in Tables A-I through A-3 of Appendix A.

2)

Uncertainty correlations are established between density and total porosity, density and effective porosity, and total porosity and effective porosity with a correlation value specified as 0.99.

3)

A random seed of 1,000 is used for the uncertainty sampling. Additionally, the Latin HyperCube Sample (LHS) method is used to generate samples of input values for the probabilistic analysis. The analysis is repeated three times, with between 300 and 500 points selected for each analysis.

4)

The RESRAD input file is then processed using the probabilistic analysis feature of RESRAD 6.1.

5)

The first 6 correlation tables of the MC

SUMMARY

.REP file are then extracted. Within these tables, the higher correlation coefficient (r2 value) between the PRCC and PCC sensitivity methods is highlighted.

6)

Sensitive input parameters is then identified and highlighted for those parameters whose sensitivity correlation value is greater than 0.25 or less than -0.25, using the correlation method identified in 5) with the highest r2 value.

7)

A copy of the RESRAD input file generated in numbers I through 3 is then made. From within this copy, the uncertainties and input distributions are deleted for all insensitive parameters (those not highlighted in step 6).

8)

Reprocessing of the RESRAD input file is then initiated using the probabilistic analysis feature of RESRAD 6.1.

9)

Once the RESRAD processing has completed the LHS step, the RESRAD analysis is be halted.

10)

The LHSBIN.DAT file is then imported into MS Excel and the MS Excel Data Analysis package is used to analyze the input parameter distributions to determine mean, 250h percentile, and 7 5 "h percentile for each sensitive input parameter.

3-10

11)

The duplicate RESRAD input file, created in step 7, is then edited within RESRAD.

Modifications include:

a.

Suppression of the uncertainty analysis.

b.

The 75th percentile value replaces the base-deterministic input value for those sensitive parameters whose coefficients of sensitivity are greater than 0.25.

c.

The 25th percentile value replaces the base-deterministic input value for those sensitive parameters whose coefficients of sensitivity are less than -0.25.

d.

The mean value calculated in step 10 replaces the base-deterministic input value for those sensitive parameters whose mean is not bounded by the 25th and 75th percentile values.

e.

Except when the coefficients of sensitivity for the distribution coefficients (Kd) are greater than 0.25, the minimum Argonne distribution coefficient (Kd) is used.

12)

The input file created in step 1 is then analyzed using RESRAD 6.1 in a deterministic mode.

13)

The 25 mrem/year dose limit is then divided by the peak dose to determine a DCGL representing exposure from all pathways. This process is performed for each nuclide, soil region, and SNEC area of concern.

14)

Steps I through 12 are then repeated with all pathways turned off, except the drinking water pathway.

15)

The 4 mrem/year drinking water dose limit is then divided by the peak dose from drinking water only to determine a DW DCGL. This process will be repeated for each nuclide, soil region, and SNEC area of concern.

3-11

4.

ANALYSIS RESULTS Radiation doses are determined for exposures to excavated and undisturbed contaminants in the Overburden soils and Bedrock for the steam plant and general site areas. The sensitivity of these projected doses to the various uncertainties of the input parameters has been examined. In accordance with the methodology presented in Chapter 3, conservative input values are selected for sensitive input parameters in order to determine DCGLs.

4.1 STEAM PLANT BACKFILL Peak doses and years of occurrence are estimated for the backfill material present in the steam plant basement. As is discussed in the methodology presented in Chapter 3, doses are estimated from exposure of an onsite resident to excavated and undisturbed materials. These doses are then compared to the 25 mrem/year limit to compute appropriate layer-and area specific DCLGs. The projected peak doses resulting from the exposure to drinking water only are also compared to the 4 mrem/year drinking water standard to estimate corresponding drinking water specific DCGLs for the steam plant backfill materials.

The resulting DCGLs are summarized in Table 4-1.

As is illustrated in the Table 4-1, DCGLs computed from exposures to excavated L

materials are most limiting when based upon the 25 mrem/year exposure standard and range from 5.6 pCi/g for Sr-90 to 41,000 pCilg for Pu-241.

DCGLs computed from drinking-water only exposures to excavated materials are significantly higher, ranging from 130 pCi/g for Pu-239 to 8.6x 102 pCi/g for Ni-63 (with specific activity as limits for Co-60 and Cs-137).

4-1 L

Table 4-1 Steam Plant Backfill DCGLs (pCi/g)

Excavated Backfiill Undistubed Backfill A t Drinking Drinking

[

All Paths Vater Only All Paths Water Only Composite 11-3 2.3E+03 2.1E+04 1.5E+02 3.5E+01 3.5E+01 C-14 4.2E+01 2.9E+04 l.lE+01 5.6E+00 5.6E+00 Co-60 8.0E+00 5.8E+06 8.OE+00 Ni-63 3.2E+03 8.6E+22

1. 1E+08 5.6E+ 18 3.2E+03 Sr-90 5.6E+00 4.9E+02 3.3E+00 1.1E+00 1.IE+00 Cs-137 2.1 E+01 5.513+08 2.IE+01 Eu-152 2.1E+01 3.8E+17 3.8E+07 9.5E+14 2.1E+01 Pu-238 1.2E+02 6.6E+03 2.913+02 9.11E+00 9.IE+00 Pu-239 1.1 E+02 1.3E+02 1.9E+00 3.OE-01 3.OE-01 Pu-241 4.11E+03 8.0E+07 4.9E+04 9.3E+04 4.IE+03 Am-241 1.1E+02 1.5E+06 1.7E+03 3.11E+03 1.IE+02 "DCGL set to individual nuclide's specific activity limit.

4-2

DCGLs computed from exposures to the undisturbed backfill materials do not follow this trend. DCGLs based on 25 mrem/year from undisturbed backfill materials range from 1.9 pCi/g for Pu-239 to 5.5x108 pCi/g for Cs-137. Drinking water only-DCGLs developed for undisturbed steam plant backfill range from 0.3 pCil/g for Pu-239 (much lower than the 25-mrem/year DCGL) to 5.6x10 8 pCi/g for Ni-63 (much higher than the 25-mrem/year DCGL).

A comparison of these four sets of DCGLs (excavated and undisturbed backfill for all pathways and for drinking water only) reveals a single set of conservative DCGLs for the backfill material (see Table 4-1). This composite set ranges from 0.3 pCi/g for Pu-239 to 4,100 pCi/g for Pu-241.

4.2 GENERAL SITE OVERBURDEN Peak doses and years of occurrence are also estimated for the subsurface overburden material generally present throughout the site area. Doses are estimated from exposure of an onsite resident to excavated and undisturbed overburden materials.

These doses are then compared to the 25 mrem/year limit to compute appropriate layer-and area-specific DCLGs. The projected peak doses resulting from the exposure to drinking water only are also compared to the 4 mrem/year drinking water standard to estimate corresponding drinking-water specific DCGLs for the site overburden materials. The resulting DCGLs are summarized in Table 4-2.

As is illustrated in the Table 4-2, DCGLs computed from exposures to excavated materials are most limiting when based upon the 25 mrem/year exposure standard and range from 5.6 pCi/g for Sr-90 to 43,000 pCi/g for Pu-241.

DCGLs computed from drinking-water only exposures to excavated materials are significantly higher, ranging from 7.4 pCi/g for Pu-239 to 7.5x 1021 pCi/g for Ni-63 (with specific activity as limits for Co-60 and Cs-137).

4-3

Table 4-2 General Site Area Overburden DCGLs (pCi/g)

Excavated Overburden I Drinking All Paths Water Only 2.2E+03 4.2E+01 8.OE+00 3.2E+03 5.6E+00 2.]E+01 2.lE+01 1.3E+02 4.6E+01 4.3E+03 1.413+03 3.7E+03 7.5E+21 4.2E+01 2.4E+16 5.7E+02 7.413+00 1.5E+06 Undistubed All Paths 1.6E+02 2.0E+00 1.6E+02 3.2E+04 2.2E+00 7.2E+02 8.4E+03 2.1 E+00 1.9E+00 1.3E+02 Am-241 1.E+02 9.813+04 1.2E+01 a DCGL set to individual nuclide's specific activity limit.

4-4 j

2.3E+00 Overburden Drinking Water Only 8.013+01 7.9E+00 6.7E+01 1.9E+04 6.OE-01 4.OE+02 1.4E+03 4.013-01 3.0E-01 2.0E+01 2.3E+00 11-3 C-14 Co-60 Ni-63 Sr-90 Cs-137 Eu-152 Pu-238 Pu-239 Pu-241 IComposite 8.OE+01 2.OE+00 8.OE+00 3.2E+03 6.OE-01 2.1E+01 2.1E+01 4.OE-01 3.OE-01 2.0E+01

DCGLs computed from exposures to the undisturbed Overburden materials do not follow this trend. DCGLs based on 25 mrem/year from undisturbed Overburden materials range from 1.9 pCilg for Pu-239 to 32,000 pCi/g for Ni-63. Drinking water only-DCGLs developed for undisturbed Overburden materials range from 0.3 pCilg for Pu-239 to 19,000 pCilg for Ni-63.

A comparison of these four sets of DCGLs (excavated and undisturbed backfill for all pathways and for drinking water only) reveals a single set of conservative DCGLs for the Overburden material (see Table 4-2). This composite set ranges from 0.3 pCi/g for Pu-239 to 3,200 pCi/g for Ni-63.

4.3 BEDROCK Peak doses and years of occurrence are also estimated for the bedrock material present Sthroughout the site area. Since it is unfeasible to assume significant excavation into the Bedrock, doses are estimated from exposure of an onsite resident to contaminated water pumped from the bedrock groundwater. These doses are then compared to the 25 mrem/year limit to compute appropriate layer-specific DCLGs.

The projected peak doses resulting from the exposure to drinking water only are also compared to the 4 mrem/year drinking water standard to estimate L

corresponding drinking-water specific DCGLs for the site Bedrock materials. The resulting L

DCGLs are summarized in Table 4-3.

As is illustrated in the Table 4-3, DCGLs computed from exposures to contaminated water from the Bedrock when based upon the 25 mrem/year exposure standard range from 1.4

__pCi/g for Sr-90 to 32,000 pCi/g for Ni-63.

DCGLs computed from drinking-water only exposures to bedrock materials range from 0.3 pCi/g for Pu-239 to 20,000 pCi/g for Ni-63.

4-5

Table 4-3 General Site Area Bedrock DCGLs (pCi/g)

Undisturbed Bedrock Drinking All Paths IWater Only 1.3E+02 3.3E+00 1.5E+02 3.2E+04 1.4E+00 6.6E+02 5.7E+03 1.8E+00 1.6E+00 8.7E+01 9.9E+00 I

3.1E+01 5.4E+00 6.7E+01 2.0E+04 6.OE-01 4.0E+02 1.4E+03 4.0E-01 3.013-01 2.013+01 2.3E+00 SComposite I

3.IE+01 3.3E+00 6.7E+01 2.OE+04 6.OE-01 4.OE+02 1.4E+03 4.OE-01 3.OE-01 2.OE+01 2.3E+00 4-6 I

H-3 C-14 Co-60 Ni-63 Sr-90 Cs-137 Eu-152 Pu-238 Pu-239 Pu-241 Am-241 L

li L

L_

L

A comparison of these two sets of DCGLs (all pathways and for drinking water only) reveals a single set of conservative DCGLs for the Bedrock material (see Table 4-3).

This composite set ranges from 0.3 pCilg for Pu-239 to 20,000 pCi/g for Ni-63.

4.4 Kj UNCERTAINTY As is described in the methodology presented in Chapter 3, the effects of distributions of uncertainties in input parameters is conservatively incorporated into the estimation of the DCGLs by using 25-percentile and 75-percentile input values for those parameters for which the resulting DCGLs are found to be highly sensitive. When available, input parameter distributions are derived from onsite measurements (e.g., water table fluctuation, individual layer thickness, etc.).

li When not available, RESRAD and DandD distribution defaults are employed.

These distributions are summarized in the Tables included in Appendix A.

A range of site-specific Kd values is included in the distributions given in Appendix A.

Individual Kd values were measured for the various materials available at the site. While the L

minimum and maximum of these Kd values is included in the distribution tables, measurement uncertainty and variation for each material type have not been examined. While a uniform profile L

is specified for the distribution shapes of H-3 and C-14 (based on historical data - Thibault 1990),

L no corresponding nuclide-specific distribution shapes can be specified for the remaining materials and isotopes. In addition to this, the use of available site materials for backfill activities as part of L

the decontamination efforts is also desired.

L_

Because of this, the sensitivity of the composite DCGLs to uniform distributions in Kds, bounded by the site materials minimum and maximum values, was also evaluated. A uniform distribution was selected in order to minimize the preferential selection of one material's Kd as part of the analysis. This selection is consistent with the RESRAD default distributions.

4-7

The RESRAD model default Kd distribution shape for nuclide Kd is generally identified as Log-Normal-N. Additionally, RESRAD includes means and geometric standard deviations as characteristic parameters for the default distributions.

Examination of these characteristic parameters reveals that the site-specific Kd ranges are far outside of the RESRAD mean peaks.

For example, the RESRAD mean Kd for Am-241 is 7.9, with a geometric standard deviation of 2.3.

In comparison, the site specific values range from 1,000 to 5,000. Examination of the section of the RESRAD default distribution shape between the values of 1,000 and 5,000 reveals a relatively uniform shape.

Examination of the effects of variations and uncertainties on the site-composite DCGLs revealed many were unaffected by Kd variation (generally those nuclides for which the groundwater pathway is not a major contributor to the overall dose). For these cases, inclusion of L

the variations in Kd did not change the identified list of input parameters for which the DCGLs

" were sensitive.

For some isotopes (particularly those for which groundwater was a major contributor to the overall dose), the Kd parameter was identified as one for which the results were sensitive. For L

these nuclides, inclusion of variations in Kd did not change the sensitivity of the resulting DCGL to variations in other previously identified parameters. Additionally, examination of the nature of relation between Kd and resulting DCGL revealed a negative correlation (e.g., increases in Kd L

result in decreases in dose and increases in DCGL). Because of this, the methodology in Chapter 3 would suggest a selection of the 25 percentile value from the distribution of Kd values would be L

conservative. Therefore, the use of the minimum Kd, which are lower than the 25% value is L

conservative and maintains the freedom of being able to use any site material in decontamination activities.

L L

4-8 L

5.

SUMMARY

AND CONCLUSIONS In preparing to terminate the NRC license for the site, GPU determined DCGLs for the top meter of the SNEC site soil that correspond to the 25 mrem/year radiation dose limit prescribed by NRC for site cleanup. This report documents a conceptual model and methodology developed by URS to determine DCGLs for the sub-surface zone below the top meter of soil.

Two areas of concern are considered for estimating radiation doses for a resident / farmer scenario: the general site area (including the CV and the Spray Pond Areas) and the Steam Plant.

Input parameters for which the resulting peak doses are most sensitive are identified.

Conservative values are selected for each of these parameters and conservative DCGLs are calculated from the resulting peak radiation doses.

The site hydrology is dominated by a shallow Fill layer and a deeper Bedrock region, separated by a relatively impermeable Overburden layer. The Bedrock drains westward toward the River. Disturbed areas (or planned disturbed areas) of the Overburden at the perimeters of the CV, Steam Plant, tunnels, and Spray Ponds hydraulically connect the Fill and Bedrock and enhance drainage from the Site.

L Radiation exposure pathways associated with the resident / farmer scenario are analyzed to estimate radiation doses. Gamma radiation exposures occur in the yard and through the house L

floor from radionuclides mixed into surface soils from excavation and well cuttings. Exposures from inhaling contaminated dust occur during site grading, well excavation, garden tillage, and L

vwind resuspension. Exposures from ingesting contaminated soil occur from soil entrained on vegetables and unwashed hands. Exposures from ingesting contaminated drinking water occur L

from transport in the Bedrock. Exposures from ingesting contaminated fruits and vegetables L

5-I

occur via their uptake from contaminated surface soil and contaminated irrigation water.

Exposures from meat and milk occur from contaminants in animal feed and water. Gamma radiation exposures occur in recreation while fishing, boating, and swimming.

Additional recreation exposures also occur from ingesting contaminated water while swimming and from consuming fish from the River.

RESRAD Version 6.1is used to estimate and combine the exposure distributions for the critical times.

RESRAD Version 6.1 computes probabilistic estimates of radiation dose distributions that result from various distributions of input parameters.

In order to account for plans of using heterogeneous Site materials for backfill and remediation, the lowest nuclide distribution coefficients are used in the analysis. This minimizes transport retardation and decay of contaminants, before they reach the point of exposure. This modeling conservatism allows single assessments of the Bedrock and Overburden layers to be conservatively applied site-wide. The first application of RESRAD represents the Fill layer and associated surface exposures.

The second evaluates ingestion of drinking water and use for irrigation from a well drilled into the Bedrock.

NRC's site cleanup criterion of 25 mrem/year and the EPA 4 mrem/year drinking water criterion are used to determine the DCGLs for each nuclide in each subsurface material layer, based on the temporal peaks of the mean doses. As are listed in Table 5-1, the most limiting DCGLs are conservatively proposed as site-wide subsurface DCGLs for the materials deeper than one meter.

This composite set ranges from 0.3 pCi/g for Pu-239 to 32,000 pCi/g for Ni-63.

5-2

Table 5-1 Site-Wide Composite Subsurface DCGLs (pCi/g)

BACKFILL COMPOSITE 3.5E+01 5.6E+00 8.0E+00 3.213+03 1.1E+00 2.1E+01 2.111+01 9.1E+00 3.OE-01 4.1 E+03 1.113+02 OVERBURDEN COMPOSITE 8.0E+01 2.0E+00 8.0E+00 3.213+03 6.013-01 2.1E+01 2.1E+01 4.OE-01 3.013-01 2.013+01 2.313+00 BEDROCK COMPOSITE 3.1E+01 3.313+00 6.7E+01 2.0E+04 6.0E-01 4.0E+02 1.413+03 4.013-01 3.0E-01 2.0E+01 2.313+00 SITE-WIDE SUBSURFACE COMPOSITE 3.IE+01 2.OE+00 8.OE+00 3.2E+03 6.OE-01 2.1E+01 2.1E+01 4.OE-01 3.OE-01 2.0E+01 2.3E+00 5-3 11-3 C-14 Co-60 Ni-63 Sr-90 Cs-137 Eu-152 Pu-238 Pu-239 Pu-241 Am-241 L

L

REFERENCES GPU GPU

Nuclear, Inc.,

Saxton Nuclear Experimental Corporation

Facility, Decommissioning Environmental Report, Revision 1, February 2, 2000.

GPU GPU Nuclear, Inc., Decontamination and Decommissioning Engineering, Saxton Nuclear Experimental Corporation, Historical Site Assessment Report, March 2000.

GPU GPU Nuclear Inc., E-mail received II January 2002 from Mr. Barry Brosey of GPU Nuclear to Dr. Kirk K. Nielson of URS, 2002.

Grove Grove Engineering. "Microshield Version 5 User's Manual.", Grove Engineering.

1996.

H&A Haley & Aldrich, Inc., Letter dated August 29, 2001 from Charles R. Butts to J.

Patrick Donnachie, GPU, Inc., Middletown, PA., 2001.

NRC U.S. Nuclear Regulatory Commission, "Update of Part 61 Impacts Analysis Methodology," 0.1. Oztunali and G.W. Roles, Washington D.C.: U.S. Nuclear Regulatory Commission report NUREG/CR-4370, January 1986.

NRC U.S. Nuclear Regulatory Commission, "Preliminary Guidelines for Evaluating Dose Assessments in Support of Decommissioning," J.W.N. Hickey, Washington D.C.:

U.S. Nuclear Regulatory Commission Memorandum dated March 16, 1999.

NRC U.S. Nuclear Regulatory Commission, Issuance of Amendment No. 11 to Amended Facility License No. DPR Saxton Nuclear Experimental Corporation, May 20, 1992.

NRC U.S. Nuclear Regulatory Commission, NMSS Decommissioning Standard Review Plan, NUREG-1727, 2000.

Palisade Palisade Corporation, @Risk Risk Analysis and Simulation Add-In for Microsoft Excel, Version 4.0, March 2000.

Thibault Thibault, D.H. et al. A Critical Compilation and Review of Default Soil Solid/Liquid Partition Coefficients, Kd, For Use In Environmental Assessments. Atomic Energy of Canada Limited report AECL-10125, March 1990.

URS URS Corporation.

"Proposed Conceptual and Calculational Modeling for Determining Sub-Surface DCGLs for the Saxton Nuclear Experimental Corporation Site." (RAE-42613-003-5030-2), August 2001.

Yu Yu, Charles. "RESRAD Version 6.1 Dose Assessment Model" Argonne National Laboratory, Environmental Assessment Division, 9700 South Cass Avenue, Argonne, IL, 60439, May 31, 2002.

R-I

L APPENDIX A RESRAD INPUT DISTRIBUTIONS L

L L

L L

L K

Appendix A This appendix contains the input parameter distnbutions that are used to determine the DCGLs for the Steam Plant Backfill, Overburden layer, and River sediments. The resulting peak doses and DCGLs for these subsurface layers are also summarized.

A-2

TABLE A-1 Steam Plan Backfill Input Paramete, Dstrbeutios In RIS

.0: Io

,u I1 In" on pI up.

R:D P

Dniy Of CoýrMfrntdZoe(/

P" DnsiyoffCover Maeil(/)

r=

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P Satrated ZoneHdalc"Gaden P

Saturated ZoneTtlPrst SauaeConemý Un-..

CZ.i1 P

DnI.- of Unatraedn Zoe..

I

~1 P Thcknes.f Unsatnrate Zone I up, P TowM op~oe, of Unsnbjte Zone "P Unsauraedone

_I b Fae=eo "P UnsaturpWe Zoe FmedC l

"P E4emn GenMa Shil.n ract.

".O lr~c ust Flffraa laop.

1.60 lot In.

0.59 S)iaI used 0.2 Ovod,..

0906 0.35 5.0Cr" 1.8 ClssPARAINETERS RESRAO input P

T11cma~ss fSu Evsin Lae Mf Cý14 in Sul (ft) 0.3 P

Ehoemoumcdsffvn Factors FreshWate Defollt -Val P

Fool Transfe Fa..r defaulIt -Val po eocnnafdn~cxQ 2000 Bp asic akn iC..ose.mfin NC2 P TZ ene =

mp. be)lhcl (VT P

Time for W.u."c (w')

Times for 10sns(T o

T...s fur.02u;on eT P

AvrgAnalWnMFed(~sc P

Cont.&nnas Zone Fie. Cammacity P Contamnat Zop. TotalPoroi 0.46 g

ooDm ootrs 2.505+00 2.05+0 030.4 lion

ýWA N/A I

A A

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fillsuda

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1.

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-s-t oe.- No-disers-icu N~o d es-e ic4(

)

T Non-0loes.io

'I S. seaed Zone bmul Pn retsiej Saturtede Zon. Sfec. Pcosit

,,satrate Zone Heldral CondR'omy(/

Sat urate Zon l-ytauioOaet

.s er TIo Dropm nat I(('d, WE=l Pump 1InelkU neim..

(.i1 Welt

~

~

~

1 Pupn Ra-nV4r

$7.91l i5.5 I

I~3

ills, I

4 30,2 solt 1

.7 t.1

0. 41

[

0.2

old, Number of Unsnltnotej Zone Sfrala I

i WA ji WA I

T.

ota l rsit of Unsaturated Zo.

I ZUn sat uraed Zoer 1 b Parameter Unatrae Zone 'Ield CDocut Eduenme Glarrma Sh'eedd, Fester Indoor Dust Fibeiaon Fadtor 11.5 5.6s I

1___

0 I 1 1

otno LI, Weillhtedid droo f~i aund oveirurdenl so Per URS Tecncl multiec 6/5/0 nod J SNEC 5/lS/O j

SNEC 5?iS SN EQ I

"/lt. 2m smexovoiu as per UAS Tellns.a Adproac 5/18,0 I~ WU" {f 1, di Cmri~f ftll 50 aeeras peir URS Toechical Aptedc~h 6/1/0 C(o3 Timms foe ODulatons Oyr Times.ce Csulatoc (v Timesifor........oc Times for Clo o isv Avofaged Miniu Wirel Spesed (W~ed, Contaminated Zcoe Field Dapacit, WA I

WA N

Rol m0om p

Rota p

kol.

p m01 p

mil p

m04 RP p

P.

P lp PK SEED ill o' YWwuxw U.I.-

TABLE A-'

General Area (herbu rden Input Paranete'r Distribudons AS NoneAsmarol S N-EC T-10102 WI.,.

.Sa0.

NES W~ 40 0~dltctt(,

in)

U_______________

NEt13 aeg 140 of*

atSa Mon Ale().

0

-0.6 TraouoNES

/30 10100 M

Woetoin Reniaai corso'ont aI

Vegaltob, 20.1

$4

riaboula, NE 510/

10100 A4 oet c, YieidtbtafyknQ NonlQAssigned SE

/00 10100 14 We-atl d,

aloO V

0.an 0.am TO~letajlr

$NEC 5/11/0 lie P

O~NES /30 O/stbunion Coefficient to, Carbon ALVle GPU M.

PUMx.

Dsrblo no0 P

SNECS/-in0 I&

P S~NECS/~0 InDisb.ut/on Co Cooen lop aba Value Use I AMt Min.

ANLMax SNn rib ut/o ISa P

SNECO 13/0 IS P

SNECS/W 3/.

I'S P

S~NEOSlS0 D/alrbufI onoCefoo.nteolrrn I0At Value PUse GPtjMh.

IP

a.

D~tbto 1W P

SNEC 5/1312 D/st,/bution CoetImeenI 'orlo,,

I au U Ie AUMm.

IP AMax itbt

,'a P

$NEC I'll.0 ti6 SNES

/00 0/s trbutlon abostt/e/nhtor 0.a Manu U.

I ANt. Min~. I Aft.bn MaD/rdnto 10 P

StIE 513/02-IS P

O~~NES

/30 t~~e P

O~NES 100 Ma Wuion.

Coabe if/c/artt Plutoniumw

~alue U

Ie AMt Min,. IAW Mon.

ifrbl, Ia P

SNCIdE 5130 IC PSdE 6/13.2

'/s prbto C o.Otatlen ro Stnll..

V nalu d

AN t. MlnI A NLM ax 0/ý/b s

2

'I S P

S N E C b I l 0 / 2 11 P

SNECS/3/0 no SNCEC$/3/

ICTE: M& d aheA.a Sybe al,r

.... van u

e I~

Ln. vatu I.'ee1 he le-ee r.,ote -va.u or this element old the ANL barx. ea-u Iho higbgqot flpotd Slu lorEaa:=

":dIem t E2y8ee raSrC.pu alua R

R N

N

TABLE A-3 Redrick ipumat Paramnfter Distributions ihnknse of Sc Evagion Layer of C-1 4 in Soil (n) loaccumulatiLn Factors. Fresh Water Food Transfer Fascors I

-

I NESRA Input Default Values LZ a ME T Values Assigled Ismalrbt~ion I.6 Triangtuar SNEC 5/13/02

_i_*

L

_*i n1N02 Area of CLaanhinoted Zone {rn2) 10r00 NA WA NWA UPS Technical Appracph 6/5/02 fEasic Radition pDose Umit (mremi (NRC) 4 N/A WA NVA SNEC 5/13/02 Lengh Parallal to Aquifer Faow (in) 112.8 NIA MA WA RESRAD Data Collection Handbook Thickness of Contarmranated Zage (in) 2.000E÷00 WA WA WA URS Technical Approach 6i5/02 Time Sir.. Placement ci Moterres sy 0

WA WA NWA SNEC 5/1 /02 Times for acultmo"s yr) 1 NWA WA WA SNEC S/i 3/02 limes for Calculatic"s "')

3 WA WA WA SNEC 5/61/02 Times for Caruulation ("r1 10 WA WA NWA SNEC 5/13/02 "Times for Calculations (yr) 35 NIA WA WA SNPC 6/13/02 Teres for Caiculatrcns (yr) 150 WA WA WA

$NEC 5/13/02 Tlmes or Calculations (yr) 30M WA WIA WA SNEC 5/13/02 Times for Calkulatlcs fyr) 1000 WA WA WA SNEC 5/13/02 Times for Caculations (yr) 10000 WA WA WA SNEC 5,13/02 Average Annual Wird Speed (rn/sec) 4.07 3.13 4,83 UnIform SNEC 5/13/02 Contaminat ed Zone Ried Capacity U8A3 0.079 0.192 Uniform SNEC 5/15/02 Contaminated Zone b Parameter 5.6 405 7.12 Uniform SNEC 513102 Contaminated Zone Eroar Rate mnkr) 0.000345 0.0000 0.0000 Lo.tunifforI SNEO 5,13/02 Contaminated Zone Hydraulic COadLctaii (mfyr) 07.91 15.59 909.52 Uifonm SNEC Sr13i02 Contaminated Zone Tootl Porosia 0.3.1 04 Underm SNEC 513/S02 SNEC 5,&1o02 A UPS Technical Cover Depth (m) 3.0 WA WA MA Approaech 6/5/02 Cover Depth Erosion Rate (rryr) 0.0003,5 0.00009 0.0006 Lzundfornt SNEC 5/1/O02 Densityo f Coarntairted Zone (gvoc)

I, 128 1.92 I

form SNEC 5/41302 (reported as rot Denity of Cover Moteal (,le.)

by RESRAD

~ 1.28 1.92 Uniform SNEC 5,13/02 Evajectranapiratlon CGoeffcient maVr) 0.50 0.5 Uniform SNEC 5/i3/02 H umcdity "

Air (gWmr3)

WA WA WA NWA UPS Techniced Approach 5/5/02 trudgaten (gyr) 0.2

-N NeMe Assirned SNEC 5/43/02 trd ortb Mode A

Overher Ovartead WA WA WA SNEC 5/4/02 Precipitation MAy) 01936 0.68 1327 Uniform SNEC 5,13/02 Runoff Coeff ieit 0.36 0.3 014 Uniform SNEC 5/13/02 Watershed Area for Nearby Streama or pond (M'S2) 5.0agEr)6 NOle Assigned SNEC 5/13/02 De-sit, of Saturated Zone (a/co) 1!6 1.28 1,2 Uniform SNEC 5,13/O2 Model: Noc-clepersice (ND) or Mass-Balance (MB)

Mass Balance WA WA N/A SNEC 5/13/02 Saturated Zone b Parameter Not Used N/A WA MA SNEC 5/1N/02 Saturated Zone Effective Porlsity 0.028 0.00M 0105 LOgulnform SNEC 5/13/02 Saturated ZOne Hydraulic Co, u.olty Qr*ny) 67.91 5.59 909.,2 Uniform SNEC 5/13/02 Saturated Zene Hydraulic Gradient 0.02 0.013 0.03 iniform SNEC 5,13/S2 Saturlaed ZoneTots] porosity 0.""

0,31 0.41 Undiform SNEOS/laI02 Water Table Drop Rate (iryr) 0 None Assijr, ed SNEC 5/13/02 SNEC 5,13/O2 10 2m rexavalOn As Well Pump Intake Depth (m) 30.2 1l2 50.2 Wniaorm per UPS Tehnical Approach 6/5/02 (reported als nm t Well Pumping Rate (rry) ud byA 207.3 365 thifnorm SNEC 5/13/02 Saturated Zone Rard Capacity 0.136 0.079 0.192 Undform SNEC 5/13/02 Density 0f Unsaturated Zone 1 (glee) 1.6 1.28 1.92 wihforhe SNEC 5/13E02 EfPfetive Porosity oe Unsaturated Zone 1 0.41 0.8 0

I54 Uniform

$NEC 6/13/02 HysdmUlic Conduam vity of Unsaturated Zone 1 (rrfyr 67.1 15.59 009.53 Louniforam SNEC 5/1 302 NMttberc Usartuat ed Zone Strat 1

WA WA IWA SNEC 5/13/02 Thikness of Unsaturated Zone 1 tu)

Total porosity of Unsaturated Zone I Unsaturated ZOne I b Parameter Unsaturated Zone Field 1aRanitr sstemial G(amma Shieling Fasctor 1ndor. Dust Ilt rami FIct, 0,ia10 N/A WA WA use I fochalmn Approeach /5/02 oeffectively zero) 0.35 0.5I Uniform SNEC 5/13/02 URS Technical Approach V05/02

,-.N i

UA1r i

MS

'.nnlcc d ZpprJ M

U201 2 CJWA 4

WA 4

WA 4.WA URS Technical Approach 6/S/OS ingism ajo f

ate For mt t)

Mass Loaofngi kr Inh*alson (&'rTWS)

Contmponatied Fracton of Drdnking Water Contarminated Fraction of Househodd Water Conterranatae Fracger c0 Irrngtios Water Conasnuated Framchon of Lvestock Water WA WA WA WA UPS Technical Approach 615/02 WA WA NIA WA URS Technical Approach 0/SO WA WA WA NA UPS Technical Approach 6N5102 1-None asusigned SNEC 5&13/02 WA N/A NIA WA URS Technical Approach &/5/02 WA WA WVA WA UPS Technical Approach 6/5/02 WA NWA WA WA URS Technical Approach 6P5/02 WA Wit WA NVA UPS Technical Acproach 6/5/02 WA WA AM WA UPS Technical Approach 6/6/02

/A WA NWA NA URS Technial Approach 6/5/02 "MA,4 1860 Truncated Loceiormst-N SNEC 5,13/OP WA WA Net URS Temhnical Approach G/5/02 WA WA NIA NIA UPS Technical Approach 6/5i02

.10 ga1a M017 1017 nala aOli vi1a

.11 fore 1010 101m p

P C. P 0, P B, P B, P U,BP M, B

Tuen.

or T.

n, Dh

" Sol MA, Ln'.,

pru) wundelm Runt"d

ýstun Ef Wl VKETAM W.

Gn), Y."d for FEMer (korr'2) t"Int cup Yl"d for Lead, (WW2)

Wet Cm Mi.

lor NýLarf, (kg(ru"2)

InrI F.W.

d etarl.

URS T-Mical AMT-l Vý01 URSTschrucW

-h Sn2 URS Turhrucal Apprc-h

ýý02 SNEGýIN02 SNEC ýIW2 SNEO ýIW02 SNECýIW02

ýNýý

ý' W2 SNEC &IX02 SNEG ýIS02

$NEC ýIX02 SNEG ýIW02 SNECý1=2 SNEC 511WO2 SNEC &1&02 S

0 SN32.

21122022 SNEC ý1XO2 SNEG 611W.

.02 S

2 SUP zVffl

$NEC ý1&02 SNEC ý1ý02 SNEC51W2 SNEG TWIS02 CE 02 SNUG NEW SNEOEI/1=2 SNEC&IN02 SNECV1=2 SNECWXW SNECV1&02 SNEC91ý02 SNEClaQ2 SNECS11ý.

SNEC&lVO2

$NEC&1102 SNEGýlM,2 SNE(r5r1&02 coý

6 p

Rl.

I IIistribufmCnefflunurtforStrontim pulwLtuutl I

ANLMin, R

Tnu p

DWrbupon Cmffwmnt W

Umurturn Value thuni 1

ANL

m.

I P

NOTE ML M Wl-MW I, 'ý-

in,,'

111-Th, ML ur.

r'..

1"..,....

MOTE Il-:l REDVý,I1I1mSNEEblP,,tlIIInE.

NOTE M

E ffWA WA I

WA WA URS TErM" AMýh 616102 A

WA I

MA I

URS T.Mral M.-h WW02 A

puA I

I WA LARS TýhruW Appýh Vý02 mut Etnlý I;ur, WIN I

I WA URS T-hulal App,,h

&5102 WA WA WA URS Tu,,MuW ftpurach Wý02 W

IN WA WIN M,

UHS I ýMual 4pur-h

&5T02 WA WA URSTurhunWItE unIchWtW02 I

p 4

1 A

LARS TýMEM Appýh

&6102 WA FLFA WA WA URS T-hnkW App-h

&ý02 WA WA I

WA I

URS T.M.W 4,-h WtV.2 MA A

I WA I

WA URS Toahnral A

I STOR B

lpmmmmm a

OEM B

TABLEA-3 Bed-k Input Paranninr DuAributianu

.I.

.In Mu.

Ru

.:E

.1.

MI.

.1.

MB

TABLE A-4 DOSE AND DCCL

SUMMARY

F -

~

~~~~

• PRA *ODADGNRLAE mPeak Dos 2.2E+03 2.78E-03 142.03 2

4.9E+01 106603X 3.76.03 8

3.2E,03 5.31E-.=

7.6E+21 4,879 5.8E+00 9.456-0 4.E.01 2.12.0 2.1E+01 1.696-IS 2.42.16 3,587 1.32.02 6.966-0 6.72.0 471 4S2.0 6.39E-01 7.4E.+00 894 432.0 2.64E46 15.06 3.949 1.12.0 4.106456 9.82.04 1,04 1.M.201 1.61 0

11,32.0 2.0 0

1.56SE41 1's 6

1106.01 2.2 0

3.E48E-7.2 9

2.972-03 8.4 4

1.182+01 2.1 1.316+01 1.9 1 SEE-O 1.3 0

2.14E400 1.2 4

E+00 2.02 E+01 pek os 0

6,94E-02 612.0 2.066-04 19.04 4

6.55F+40 GOCE-O 0

1.016-0 4.06,02 5

2.77E403 1.460 2

9.492E40 CO4.0 1.076+01 302-Cl 1.73E+0 2.32.00 7

I Excavated Backfill on Surface Peak Don (mrem*yr 1.002-02 23!

0 6,94E-0 4.2 0

3.092+0 8.0!

0 7.766-03 3.21 0

4.46+0 6.6!

0 1,182.0 2.11 0

112.00E 2.1!

0 2.022-01 1.2!

0 2.ME4-C 1.1!

0 6.14E.03 4.1!

m8 2.322-0 1.1!

0 E+03 E+01 E400 E+01 E.01 E.0 I [mremy 1.4GE-04 2.96.04 9

4.!64E-2

!.6M,=

4,881 8.1m.03 cm9+0 41 1.04E-17 386.17 3,131 6.076-0 6.66.03 471 3.13E-02 1.3E+02 7M6 4.97E468 8&O6,0 1,60 2.74E-0 1.56.06 4.455 I

Undistubed Backfill 1222Ww" 4,61 7.450 Do

.36M 0 3t.39,0 Do.

+

I.612-02 tE&02 4.38641 9.16+0 1

31 5.02-04 4.96*04 4312.05l 9.36.

1,M9 4,461 1.48602 1.7M.0031.39E-00 316+03 1,702 1,48 2.0E+.00 StOE-C 2.12,01 2.12.01 4.0641

&OE-0l 2.02.01 22E6+00 nbaedl on 4 nlrerTyr)

Lim It-)

H43 C-i14 Eu-152 Pu -24 1 Peek Dose Qn..dlr) 1.146-02 0

6.94241 0

3.N9E.00 0

7. 76E4 0 0

4AO4020 0

1.18E+0 0

1.2 1E.0 0

2.006-01 0

5. 4 2 01 970 8.86E2-0 34 2.322-C 0

SSGS I

I BEDROCK I

Sl TE SU R FACE IB' ISNU

",L,ý,

DCGL