ML24303A282

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License Termination Plan Supplemental Submittal for Technical Bases Documents
ML24303A282
Person / Time
Site: Oyster Creek
Issue date: 10/29/2024
From: Noval W
Holtec Decommissioning International
To:
Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation, Document Control Desk
References
HDI-OC-24-040
Download: ML24303A282 (1)
v • d • e


Text

Krishna P. Singh Technology Campus, 1 Holtec Blvd., Camden, NJ 08104 Telephone (856) 797-0900 Fax (856) 797-0909

10 CFR 50.90 10 CFR 50.91 October 29, 2024

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR -16 Docket Nos. 50-219 and 72-15

Subject:

Oyster Creek License Termination Plan Supplemental Submittal for Technical Bases Documents

Reference:

(1) Letter to United States Nuclear Regulatory Commission (USNRC) from Holtec Decommissioning International, LLC (HDI) titled License Amendment Request to Revise Oyster Creek Nuclear Generating Station Renewed Facility Operating License to Add License Condition 2.C.(18) to Include License Termination Plan Requirements, ADAMS Accession Number ML24214A037

Pursuant to the submittal of Reference (1), HDI is submitting the Technical Bases Documents (TBD) as required by the USNRC to support the review of the Oyster Creek License Amendment Request. Enclosures (1) and (2) are attached to this letter and provide the documented deliverables in Reference (1).

This letter contains no new regulatory commitments.

If you have any questions or need further information, please contact Steven Johnston, Oyster Creek Regulatory Assurance Manager, at (609) 971-4325.

Sincerely,

William Noval Director, Regulatory Affairs Holtec Decommissioning International, LLC

HDI-OC-24-040 Page 1 of 2 Krishna P. Singh Technology Campus, 1 Holtec Blvd., Camden, NJ 08104 Telephone (856) 797-0900 Fax (856) 797-0909

Enclosures:

1). In Situ Object Counting System (ISOCS) as Applied to Final Status Survey at Oyster Creek Nuclear Generating Station (OCNGS) 2). Method for Determining Nuclide Fractions and Gross Activity DCGLs

cc: USNRC Regional Administrator, Region I USNRC Project Manager, NMSS - Oyster Creek Nuclear Generating Station USNRC Region I, Lead Inspector - Oyster Creek Nuclear Generating Station Assistant Commissioner, Air Quality, Energy and Sustainability, NJ DEP Assistant Director Radiation Protection Element, NJ Bureau of Nuclear Engineering

HDI-OC-24-040 Page 2 of 2 Enclosure 1 to HDI-OC-24-040

In Situ Object Counting System (ISOCS) as Applied to Final Status Survey at Oyster Creek Nuclear Generating Station (OCNGS)

Table of Contents

1 Introduction............................................................................................................... 4 2 Equipment and Materials.......................................................................................... 4 3 Methods.................................................................................................................... 5 3.1 Defining a survey unit....................................................................................... 5 3.2 Class 1 Survey Units......................................................................................... 5 3.3 Class 2 Survey Units......................................................................................... 6 3.4 Class 3 Survey Units......................................................................................... 6 3.5 Concrete Surfaces............................................................................................ 6 3.6 Excavations....................................................................................................... 7 3.7 Efficiency Calibration for the ISOCS System.................................................... 7 3.8 Designing Survey Unit Scans............................................................................ 8 3.9 Documentation of Surveys.............................................................................. 10 4 Considerations for Use of ISOCS for FSS Applications.......................................... 10 5 Conclusions............................................................................................................ 11 6 References............................................................................................................. 11 7 Attachment 1: Demonstration of Achieving Required Detection Limits for Soils in a 10- minute Count Interval........................................................................................ 12 8 Attachment 2: Be5030V ISOCS Detector Characterization.................................... 15

Figures Figure 1 Mirion BEGe Broad Energy Detector................................................................ 4 Figure 2 ISOCS Orientation to a Wall Surface................................................................ 7 Figure 3 Geometry of a Scan.......................................................................................... 8 Figure 4 VSP Plotted ISOCS Scan Locations................................................................. 9 Figure 5 An Array of 2 m high ISOCS Sans for 100% Coverage.................................. 10 Figure 6. Side-on view of geometry. The soil disc is brown, detector crystal purple, and shielding grey. The aluminum window, due to its thinness, is not visible...................... 13 Figure 7. Source particle distribution............................................................................. 14

Tables Table 1 Scan Survey Coverage Requirements............................................................... 5

2 l Page

Table 2 Soil Values......................................................................................................... 6 Table 3 Surface Values................................................................................................... 7 Table 4. Simulation output............................................................................................. 14

3 l Page 1 Introduction

This report presents the use of in situ object counting systems (ISOCS) using a Broad Energy Germanium (BEGe) detector and associated equipment purchased specifically for performing measurements at Oyster Creek Nuclear Generating Station (OCNGS) to demonstrate that the release criteria established by the Nuclear Regulatory Commission (NRC) or other entities have been met for open land areas and surfaces.

Radiation surveys are a crucial component of final status surveys, particularly in the context of decommissioning nuclear facilities and ensuring environmental safety. These surveys provide essential data on the levels of residual radioactivity, helping to confirm that sites meet regulatory requirements for safe reuse or release.

Traditional gamma scan surveys have been performed with handheld instruments such as Nal(TI) scintillation detectors for the detection of gammaemitting radionuclides in -

potentially contaminated media or gas flow proportional detectors when alpha/beta-emitting radionuclides are potentially present on surfaces. Whichever detection method is used, it must be sufficient to demonstrate that established release criteria have been met. Data collected for this purpose must be accurate and reliable, and measurement uncertainties must be adequately addressed.

The following sections describe the BEGe ISOCS system, the scanning methods used, and their application of the output of ISOCS measurements to defined acceptance criteria.

2 Equipment and Materials

An ISOCS BEGe detector, manufactured by Mirion Technologies, has been procured for radiological measurements at the ONCGS. Other ISOCS detectors (e.g., reverse electrode coaxial detectors) may also be employed based on specific survey needs using the methods described below.

Figure 1 Mirion BEGe Broad Energy Detector

The detector system purchased includes the BEGe detector with the InSpector Multi-Channel Analyzer (MCA) unit connected to a laptop computer running the data acquisition software (Genie-2000). In some settings, the detector may be used in an open configuration or inserted into a steel or lead shield collimator. These components can be mounted on a stationary frame, on a wheeled cart, or on a platform capable of being

4 l Page

suspended by a crane. The method of deployment selected will be based on the environment in which it will be used. A battery for field operations can power all th e operations. Additionally, a wireless network may be installed, allowing the laptop running the system to be controlled remotely from another laptop, thus eliminating the need for a direct cable connection between the operator's station and the ISOCS unit.

3 Methods

3.1 Defining a survey unit

Surveys using the ISOCS are performed on survey units. MARS SIM defines survey units according to their potential for elevated radioactivity, recognizing that not all areas have the same potential for contamination. Areas with a lower potential for contamination do not require the same level of survey coverage to achieve an acceptable level of confidence that the site meets the established released criteria. To operationalize this, survey units are divided into classes : Class 1, Class 2, and Class 3. The scan unit survey coverage requirements are presented in Table 1.

Table 1 Scan Survey Coverage Requirements Classification Class 1 Class 2 Class 3 Scan Coverage 100% 1 to 100%* Judgmental (1 to 10%)

Gcor=Class=O=pvey=rtsI=the=scan=covwill==oporti=tthe=potential=f=detti ed=activity==s=cltthe=reascriterionI=per= jppfj=ction=RKRKPK=eistorical imioand=indivi=measent=rts=clted=durinchactizion=will=be=ed=to crtivity=potei=tscan=clevsK

3.2 Class 1 Survey Units

A Class 1 Survey Unit under MARSSIM refers to a designated area within a contaminated site that is considered to have the highest potential for residual radioactive contamination, typically including areas with known prior spills, leaks, or disposal activities, and therefore requires the most rigorous survey methodology to assess its radiological status accurately; essentially, it is the highest priority area for detailed radiological investigation due to the likelihood of elevated contamination levels. The primary assumption is that a potential one square meter of elevated radioactivity exists at the edge of the area being evaluated by a single in-situ measurement. The guideline associated with this scan is the Derived Concentration Guideline Level (DCGLW), where the W subscript signifies a value applied to a wide area scan. To determine the required scan Minimum Detectable Concentration (MDC), one must first determine the Elevated Measurement Concentration for a DCGL (DCGLEMC) for this hypothetical case. Table 2 presents the DCGLEMC scan requirement for Class I Areas for soil for the primary nuclides of concern at ONCGS, i.e., Cs-137 and Co-60.

5 l Page Table 2 Soil Values Nuclide DCGLw (pCi/g) Area Factor (1m2) DCGLEMC (pCi/g)

Cs-137 3.58E+01 9 3.20E+02 Co-60 7.94E+00 9 7.15E+01

3.3 Class 2 Survey Units

A Class 2 Survey Unit under MARSSIM refers to a designated area being surveyed for radioactive contamination where the expected level of residual radioactivity is below the derived concentration guideline level (DCGL W), meaning contamination is not anticipated to exceed acceptable limits but still requires a more thorough survey than a Class 3 unit with a lower probability of contamination; typically encompassing areas with a relatively uniform distribution of contamination. The investigation Level for Class 2 Survey units is presented in Table 2 as >DCGLw or >DCGLSCAN if DCGLSCAN is greater than DCGLW. The detection limits are set such that DCGLW may be detected in a 1m2 area at the edge of the field of view.

3.4 Class 3 Survey Units

A Class 3 Survey unit under MARSSIM is a designated area that has been preliminarily assessed as having a low probability of containing areas with residual radioactivity, meaning it is considered to have a low risk of contamination and requires less intensive radiological surveys compared to Class 1 and Class 2 areas. The investigation level for Class 3 Survey units is provided in Table 2 as detectable greater than background. This requirement is met by investigating any scan that positively detects activity more than the site-assessed surface soil background, if known, Cs-137.

3.5 Concrete Surfaces

As mentioned in the introduction of this TSD, the ISOCS unit may be used to perform assays of concrete surfaces (e.g., building surfaces). To perform these surveys, the ISOCS unit is positioned perpendicular to the surface using a 90-degree collimator, as shown in Figure 2. A model is developed as appropriate for the survey plan data quality objectives (DQOs) using the geometry composer within the ISOCS software. The investigation levels for surface surveys are included in Table 3.

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Figure 2 ISOCS Orientation to a Wall Surface

Table 3 Surface Values Nuclide DCGLw Area Factor (1m2) DCGLEMC (dpm/100cm2) (dpm/100cm2)

Cs-137 4.70E+04 12 5.64E+05 Co-60 1.35E+04 11 1.49E+05

3.6 Excavations

Personnel may not physically enter a survey area due to safety concerns, which usually is the case for excavations. While sampling may be performed remotely, scanning presents a challenge. ISOCS may be used to perform scans in these areas. The ISOCS modeling is developed in accordance with the DQOs of the survey plan. The detector must be positioned perpendicular to the scanned surface using a 90- degree collimator.

Depending on the depth and extent of the excavation, using an ISOCS system can be challenging.

3.7 Efficiency Calibration for the ISOCS System

The central feature of portable ISOCS technology is to support insitu gamma -

spectroscopy by applyin g mathematically derived efficiency calibrations. Due to the nature of the environment and surfaces being evaluated (assayed), input parameters for the ISOCS efficiency calibrations will be reviewed case-by-case to ensure the applicability of the resultant efficiency. Material densities applied to efficiency calibrations will be documented. In practice, a single efficiency calibration file may be applied to most tile measurements. The geometry most generally employed will be a circular plane assuming uniformly distributed activity. Efficiency calibrations will address a depth of 15cm for soil and a depth of up to 5cm for concrete surfaces to account for activity embedded in cracks, etc. Other geometries (e.g., exponential circular plane, rectangular plane, etc.) will be applied if warranted by the physical attributes of the area or surface being evaluated.

Efficiency calibrations are developed by radiological engineers who have received training with respect to the ISOCS software. Efficiency calibrations will be documented following OCNGS procedures. Attenuation by standing water will need to be modeled into

7 l Page the assay and carefully verified so as not to understate the depth. Experience has shown that it becomes difficult to meet detection limits with more than two inches of water in a scan assay. If there is evidence of standing water (i.e., a water sheen is visible on the soil media), then an appropriately thick layer of water attenuator will be added to the model.

3.8 Designing Survey Unit Scans

Surveys to address the Class 1, 2, or 3 criteria discussed above can be designed using software specific for this purpose, such as the Visual Sample Plan (VSP) developed by Pacific Northwest National Laboratory. Visual Sample Plan (VSP) supports the development of a defensible sampling plan based on statistical sampling theory and the statistical analysis of sample results to support confident decision-making. This software helps answer the questions of how many samples are needed, where the samples shoul d be taken, and others. Surveys designed in VSP typically use a triangular grid pattern to identify measurement locations. Our application assumes that the 90- degree collimators are installed and that the detector face is orientated downward and lifted to the desired height, h, above the horizontal plane. The detector's field of view is described by a circle of radius h (i.e., the base of a right circular cone, r = h).

P

h

Ca r

Figure 3 Geometry of a Scan

The detector response at point P with an energy E is a function of the concentration of the radionuclide (Ca), in Ci/m2, the calibration factor for the detector for energy (E), the height of the detector above the surface (h), in meters, and the radius of the area (r), in m2, and is described as:

2 + 2 )

() = ( )(2 1

The software VSP is used to design a triangular grid pattern for a given survey unit. A random starting point within the survey unit is selected, and the statistically determined

8 l Page number of desired samples (N) is used to establish a triangular grid with a random start point of appropriate grid spacing. This input may be determined using the familiar triangular grid equation from the MARSSIM, Chapter 5. The number of calculated survey locations (N) and the area (A) of the survey unit can be used to calculate the distance between survey points.

= 0.866 Equation 2

It is recommended that N+1 be input into the VSP calculation to ensure a slight additional overlap for a conservative margin. As shown in Figures 4 and 5, once the triangular grid is constructed, it is necessary to verify that 100% of the area is covered by plotting the scan measurement locations and the field of view for each measurement in a drafting program such as AutoCAD. Additional scan assays on the periphery of the survey unit may be readily added by plotting additional measurement locations using the sample points along the triangular grid. The manually added scan assay locations are shown in red.

Figure 4 VSP Plotted ISOCS Scan Locations

9 l Page Figure 5 An Array of 2 m high ISOCS Sans for 100% Coverage

3.9 Documentation of Surveys

Personnel specifically trained to operate the system perform data collection activities.

Data collection activities address environmental conditions that may impact soil moisture content. Logs are maintained to provide a mechanism to annotate such conditions and ensure that efficiency calibration files address the in-situ condition(s). In extreme cases (e.g., standing water, etc.), specific conditions are addressed to ensure that analysis results reflect the conditions. As previously discussed with respect to water, when unique environmental conditions exist that may impact analysis results, conservative compensatory factors can be applied to the analysis of the data.

4 Considerations for Use of ISOCS for FSS Applications

The in-situ techniques should be limited to characterized HPGe detectors using appropriate geometries and will be used in conjunction with the Mirion GenieTM software suite. All operations should be conducted in accordance with applicable site procedures.

Additionally, the following conditions must be satisfied: The geometries must be reviewed by a Subject Matter Expert (SME) to ensure they are correctly developed and accurate or conservative approximations of the media are being measured.

Caution must be used when applying geometries for ISOCS scanning. Careful verification that the environmental conditions and geometric arrangement are appropriate to the detector geometry is pivotal to ensuring the accuracy of the results.

10 l Page Field conditions may also significantly influence the practical applicability of the ISOCS as a field instrument. Experience has shown that the impact of attenuation from standing water may be particularly problematic in achieving the required detection sensitivity.

Consequently, it is recommended that standing water be avoided to the extent practical and sufficient counting times are planned for where it is impractical to eliminate.

For FSS applications, the ISOCS unit would be set up as previously described. We consider the activity of Cs-137 that would be within a scan area that is equivalent to 50%

of the DCGLW using the following input parameters:

In-situ soil density: 1.51 gm/cm3 DCGLW: 35.8 pCi/gm Scan area radius: 200 cm Scan area thickness: 15 cm calculates the ISOCS response to the DCGLW of 35.8pCi/g and yields a detector response of 13700 counts per hour. The Cs-137 activity within the scan area equal to 50% of the DCGLW is calculated as 5.095 x 107 pCi. OpenMC was used to calculate the BeGe detector response at a height of 200 cm above the scan area for this Cs-137 activity. The calculated detector response yielded a nominal 100 counts per minute above the background. This count rate should be distinguishable from background radiation for using longer count times, e.g., 10 or 20 minutes.

5 Conclusions

This report demonstrates that the ISOCS system can perform scan surveys that achieve the areal extent requirement and the detection levels needed to demonstrate compliance with regulatory levels.

6 References

1. NUREG-1575, MARSSIM "Multi-Agency Radiation Site Survey and Investigation Manual," Rev 1, August 2000.
2. OCNGS Site-Specific License Termination Plan
3. OCNGS ENG-OCS- 008 Site-specific " Derived Concentration Guideline Levels for Soil-Industrial Use Scenario"
4. OCNGS ENG-OCS- 004 Site-specific " RESRAD Building Surface DCGL Values Oyster Creek Station."
5. OCNGS ENG-OCS- 009 Site-specific " Area Factors for Soil - Industrial Use Scenario."
6. OCNGS ENG-OCS- 006 Site-specific " Area Factors for Use with Oyster Creek DCGL Values for Buildings/Structures ",
7. Broad Energy Germanium (BEGe) detector Characterization Report

11 l Page 7 Attachment 1: Demonstration of Achieving Required Detection Limits for Soils in a 10-minute Count Interval

Assumptions The following modeling assumptions were made:

  • The shielding around the detector and window assembly has a thickness of 1 detector diameter and height of 2 detector heights.
  • The soil was modeled as #105: Earth, U.S. Average, from PNNL-15870 Rev. 1.
  • The activity was simulated as monoenergetic 661.7 keV and scaled to the intensity of the gamma (intensity=0.946*0.851, equal to the chance for Cs137 to B-decay to the 11/2-level of Ba137 multiplied by the chance of the gamma itself)
  • For the detector assembly, only the aluminum window and crystal were modeled.
  • The height of the detector from the soil surface is 2 meters.

Materials The following materials from the Compendium of Material Composition Data for Radiation Transport Modeling, Report number PNNL-15870 Rev1 (Reference 3) were used:

  • #4, Air, dry, near sea level o Used for empty space inside simulation boundaries
  • #105, Earth, U.S. Average o Used for soil disc
  • #138, High-Purity Germanium o Used for detector crystal
  • #171, Lead o Used for detector shield

Approach OpenMC v.0.15.0 was used.

To create the geometry, the soil disc was first created with its center at the origin. Then all other geometries were placed relative to the disc, onaxis. To encompass the -

simulation, a large (5m) sphere was created to serve as the out-of-bounds boundary centered at the origin.

To simulate an equivalent source, the soils mass was found, multiplied by the appropriate pCi/g, and scaled to Bq to obtain the activity. This activity was uniformly distributed over the soil disc region. Tallies were energy -filtered to be within +/-1% of the 661.7 keV gamma

12 l Page and cell-filtered by the detector crystal. The activity was pre-accounted for in the source parameters; thus, the base output was scaled by 3600 to obtain a per-hour value. Each simulation used 10 batches of 10M particles.

Figures

BeGe Detector in collimator

200 cm

200 cm

15 cm (Soil volume radius 200 cm, depth 15 cm)

Figure 6. Side-on view of geometry. The soil disc is brown, the detector crystal purple, and the shielding grey. The aluminum window, due to its thinness, is not visible.

13 l Page Figure 7. Source particle distribution

Output Scenario Soil Cs-137 Contamination Counts/ hr (pCi/g)

Resident Farmer 9.6 3670+/-240 240 Industrial Use 35.8 13700+/-900 900 Table 4. Simulation output.

14 l Page 8 Attachment 2: Be5030V ISOCS Detector Characterization

15 l Page Enclosure 2 to HDI-OC-24-040

Method for Determining Nuclide Fractions and Gross Activity DCGLs Method for Determining Nuclide Fractions and Gross Activity DCGLs

Table of Contents

1 Executive Summary.................................................................................................. 3 2 Introduction............................................................................................................... 4 3 Calculation Methodology.......................................................................................... 4 4 Calculations.............................................................................................................. 5 4.1 Reactor Building Basement Concrete Results.................................................. 5 4.2 Turbine Building Basement Concrete Results................................................... 7 4.3 New Radwaste Building Tunnel Concrete Results............................................ 9 4.4 Gross Alpha DCGL......................................................................................... 11 4.5 Modification of DCGLGA values....................................................................... 11 5 Assumptions........................................................................................................... 11 6 Conclusions............................................................................................................ 11 7 References............................................................................................................. 12

List of Tables Table 1: OCNGS Building Surface DCGLs..................................................................... 4 Table 2: OCNGS Reactor Building Concrete Assessment.............................................. 6 Table 3: OCNGS Turbine Building Concrete Assessment.............................................. 8 Table 4: OCNGS New Radwaste Tunnel Concrete Assessment.................................. 10

2 l Page 1 Executive Summary

This document outlines the methodology for calculating Gross Activity Derived Concentration Guideline Levels (DCGL GA) for decommissioning the Oyster Creek Nuclear Generating Station (OCNGS). These DCGLGA values are essential for ensuring the site complies with the Nuclear Regulatory Commission (NRC) and the State of New Jerseys release criteria for radiological safety. The DCGLGA values are based on site-specific radionuclide mixtures, focused on the predominant plant-related gamma-emitting radionuclides Cs-137 and Co-60.

The calculations involve a multi-step process, using nuclide fractions obtained from radionuclide-specific analyses of OCNGS structures to establish DCGLGA values. The methodology applies the unity rule, normalizing the contributions of individual radionuclides to their respective concentration guideline levels. This ensures that surface contamination assessments for decommissioned structures are conservative.

To derive these radionuclide fractions, comprehensive sampling and analysis were conducted on various OCNGS structures, including the Reactor Building, Turbine Building, and New Radwaste Tunnel. The results show variability in radionuclide concentrations due to site-specific conditions such as operational history, water ingress, and other environmental factors. The DCGLGA values are adjusted to reflect these differences in contamination profiles.

In the absence of specific, site-relevant alpha-emitting radionuclide data, a precautionary and conservative approach is employed by using Am-241 as the basis for alpha contamination assessments. This ensures compliance with regulatory standards while allowing for further data that will emerge from ongoing characterization. The document also outlines a process for reassessing DCGL GA values during the Data Quality Objective (DQO) phase, ensuring that any potential changes in radionuclide distributions or contamination levels are appropriately managed as decommissioning progresses.

This methodology and its conservative assumptions ensure that the decommissioning of OCNGS adheres to regulatory standards, providing a clear path for safely releasing the site for future use.

3 l Page

2 Introduction

Evaluating the compliance of soils at the Oyster Creek Nuclear Generating Station (OCNGS) with the established release criteria is a relatively straightforward task that entails a radionuclide-specific analysis. Individual survey plans utilize the unity rule, wherein each nuclide is compared to its corresponding Derived Concentration Guideline Level (DCGL). However, when assessing building surfaces, the portable rate meter instruments employed at OCNGS, such as the Ludlum 3001 in conjunction with GM, scintillation, or proportional detectors, yield only gross activity readings rather than nuclide-specific measurements. In such instances, it is necessary to calculate a gross activity DCGL (DCGLGA) based on a specific radionuclide mixture pertinent to the site.

The DCGL GA should be set to a value considering the plant -related radionuclide fractions.

Calculating the DCGL GA value involves evaluating the nuclide fractions present and using the process defined in this Technical Support Document (TSD) to determine suitable concentration guideline levels.

Due to the variability associated with the OCNGS structures and the possibility of changes in this variability during the demolition and decommissioning processes, the individual survey unit DCGLGA values will be established during the Data Quality Objective (DQO) process. This will occur while developing the survey plans for each specific survey unit.

3 Calculation Methodology

The OCNGS site-specific Occupancy DCGLs for building surfaces have been developed and are provided in Table 1. The 25 mrem/y DCGLs are aligned with NRC release criteria, providing a standard for safe public exposure post-decommissioning.

The 15 mrem/y values represent the more conservative State of New Jersey release criteria, ensuring compliance with local regulations.

Table 1: OCNGS Building Surface DCGLs 25 mrem/y 15 mrem/y 25 rem/y 15 mrem/y Nuclide DCGLa DCGLb Nuclide DCGLa DCGLb (dpm/100cm2) (dpm/100cm2) (dpm/100cm2) (dpm/100cm2)

Am-241 1.65E+03 9.90E+02 Nb-94 1.85E+04 1.11E+04 C-14 6.42E+06 3.85E+06 Ni-63 1.63E+07 9.78E+06 Cm-243 2.40E+03 1.44E+03 Np-237 1.34E+03 8.04E+02 Cm-244 3.03E+03 1.82E+03 Pu-238 1.88E+03 1.13E+03 Co-60 1.35E+04 8.10E+03 Pu-239 1.70E+03 1.02E+03 Cs-137 4.70E+04 2.82E+04 Pu-240 1.70E+03 1.02E+03 Eu-152 2.71E+04 1.63E+04 Pu-241 6.75E+04 4.05E+04 Eu-154 2.56E+04 1.54E+04 Sb-125 7.52E+04 4.51E+04 Fe-55 3.85E+07 2.31E+07 Sr-90 8.14E+04 4.88E+04 H-3 2.10E+08 1.26E+08 Tc-99 5.05E+06 3.03E+06

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25 mrem/y 15 mrem/y 25 rem/y 15 mrem/y Nuclide DCGLa DCGLb Nuclide DCGLa DCGLb (dpm/100cm2) (dpm/100cm2) (dpm/100cm2) (dpm/100cm2)

Mn-54 5.34E+04 3.20E+04

a The 25 mrem/y values represent values corresponding to the NRC release criteria.

b The 15 mrem/y values correspond to the State of NJ release criteria.

The process for calculating DCGLGA values consists of the following steps:

1. Determine the mean relative fraction (f) of the total activity contributed by the plant-related radionuclide based on radionuclide-specific analyses. The relative fraction is derived by dividing the nuclide-specific mean activity concentration by the mean total activity present.
2. Obtain the DCGL value for each plant-related radionuclide identified in the analyses. DCGL values are provided in Table 1.
3. Substitute the values of f and DCGL in the equation below.

Equation 1: Calculation of Gross Activity DCGL

= 1 1 + 2 +

1 2 Where:

f1, f2,, fn are the mean fractions of the radionuclides DCGL1, DCGL2,, DCGL n are the nuclide-specific Derived Concentration Guideline Levels 4 Calculations

The following calculations present the methodology for developing the DCGLGA values for the OCNGS structures. Individual Survey Unit DCGLGA values may be modified, as applicable, to account for the actual characterization values for the specific area being surveyed. Note: the DCGLGA is not fixed and as more remediation or characterization data is obtained from individual survey units then the DCGLGA may change.

4.1 Reactor Building Basement Concrete Results From August 9, 2023, to August 23, 2023, a comprehensive characterization study of the OCNGS Reactor Building basement concrete was undertaken. The study involved collecting fifty biased samples from specific locations on wall and floor surfaces. The sam ples were analyzed using gamma spectroscopy methods. Cs -

137 and Co -60 were the only gamma-emitting plant -related radionuclides that were positively identified. Table 2 provides the analytical results of this effort.

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Table 2: OCNGS Reactor Building Concrete Assessment Location Cs-137 % Cs-137 Co-60 % Co-60 (pCi/g) (pCi/g)

RCA-RXB-001-B Wall 1.33E+00 51.8 1.24E+00 48.2 RCA-RXB-002-B Floor 7.30E+03 98.9 8.05E+01 1.1 RCA-RXB-003-B Floor 3.40E+02 81.5 7.70E+01 18.5 RCA-RXB-004-B Wall 8.63E+01 95.0 4.55E+00 5.0 RCA-RXB-005-B Wall 2.72E+01 88.3 3.60E+00 11.7 RCA-RXB-006-B Wall 1.47E-01 36.8 2.52E-01 63.2 RCA-RXB-007-B Wall 3.29E+00 81.6 7.42E-01 18.4 RCA-RXB-008-B Floor 7.51E+02 97.3 2.11E+01 2.7 RCA-RXB-009-B Floor 1.69E+02 99.3 1.26E+00 0.7 RCA-RXB-010-B Wall 8.47E-01 72.8 3.17E-01 27.2 RCA-RXB-011-B Wall 4.77E+01 99.3 3.38E-01 0.7 RCA-RXB-012-B Wall 1.08E+03 100.0 2.23E-01 0.0 RCA-RXB-013-B Floor 6.40E+00 88.9 8.03E-01 11.1 RCA-RXB-014-B Wall 4.56E+02 98.5 7.07E+00 1.5 RCA-RXB-015-B Wall 8.61E-01 86.1 1.39E-01 13.9 RCA-RXB-016-B Wall 8.03E-01 88.8 1.01E-01 11.2 RCA-RXB-017-B Wall 2.59E-01 62.9 1.53E-01 37.1 RCA-RXB-018-B Wall 3.92E-01 74.5 1.34E-01 25.5 RCA-RXB-019-B Floor 1.87E+02 98.8 2.19E+00 1.2 RCA-RXB-020-B Floor 5.19E+01 97.7 1.20E+00 2.3 RCA-RXB-021-B Wall 4.59E-01 76.1 1.44E-01 23.9 RCA-RXB-022-B Wall 4.34E+00 93.3 3.11E-01 6.7 RCA-RXB-023-B Floor 6.76E+02 99.9 6.42E-01 0.1 RCA-RXB-024-B Floor 1.14E+03 95.1 5.88E+01 4.9 RCA-RXB-025-B Floor 7.58E+02 91.6 6.91E+01 8.4 RCA-RXB-026-B Floor 4.65E+02 91.7 4.20E+01 8.3 RCA-RXB-027-B Floor 2.87E+02 72.7 1.08E+02 27.3 RCA-RXB-028-B Floor 3.13E+02 77.8 8.93E+01 22.2 RCA-RXB-029-B Floor 4.23E+02 77.2 1.25E+02 22.8 RCA-RXB-030-B Floor 1.14E+03 97.5 2.92E+01 2.5 RCA-RXB-031-B Wall 3.77E+00 17.4 1.79E+01 82.6 RCA-RXB-032-B Floor 3.72E+03 98.2 6.78E+01 1.8 RCA-RXB-033-B Wall 2.04E+01 89.2 2.46E+00 10.8 RCA-RXB-034-B Floor 6.61E+03 98.6 9.12E+01 1.4 RCA-RXB-035-B Floor 6.16E+02 79.9 1.55E+02 20.1 RCA-RXB-036-B Floor 1.03E+03 79.6 2.64E+02 20.4 RCA-RXB-037-B Wall 6.28E+01 64.5 3.45E+01 35.5 RCA-RXB-038-B Floor 4.29E+02 82.7 8.95E+01 17.3 RCA-RXB-039-B Wall 2.16E+01 94.2 1.32E+00 5.8 RCA-RXB-040-B Floor 8.49E+02 99.3 6.33E+00 0.7 RCA-RXB-041-B Floor 3.23E+02 96.5 1.18E+01 3.5 RCA-RXB-042-B Floor 3.40E+02 97.7 7.95E+00 2.3 RCA-RXB-043-B Floor 8.93E+02 93.0 6.72E+01 7.0 RCA-RXB-044-B Wall 1.50E+00 67.3 7.30E-01 32.7 RCA-RXB-045-B Floor 1.98E+03 99.2 1.65E+01 0.8

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Location Cs-137 % Cs-137 Co-60 % Co-60 (pCi/g) (pCi/g)

RCA-RXB-046-B Floor 2.39E+03 98.7 3.07E+01 1.3 RCA-RXB-047-B Wall 6.66E+00 74.5 2.28E+00 25.5 RCA-RXB-048-B Floor 1.24E+03 95.0 6.55E+01 5.0 RCA-RXB-049-B Wall 1.26E+02 96.7 4.28E+00 3.3 RCA-RXB-050-B Wall 1.16E+01 57.5 8.59E+00 42.5 Mean 85.0 Mean 15.0

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= 10.8515 4.70E + 04+ 0.1.35E + 04

= 3.43E + 04 dpm100cm2

Where:

0.85 is the mean fraction for Cs-137 4.70E+04 is the DCGL for Cs-1 37 0.15 is the mean fraction for Co-60 1.35E+04 is the DCGL for Co-60 4.2 Turbine Building Basement Concrete Results The OCNGS Turbine Building Basement Concrete was characterized from 07/20/2023 to 11/14/2023. Forty-five samples were taken from biased wall and floor surface area locations. Cs -137 and Co-60 were the only plant -related gamma radionuclides positively identified during this characterization. Table 3 provides the gamma analytical results of this effort.

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Table 3: OCNGS Turbine Building Concrete Assessment Location Cs-137 % Cs-137 Co-60 % Co-60 (pCi/g) (pCi/g)

RCA-TURB-001-B Wall 2.91E-01 86.4 4.57E-02 13.6 RCA-TURB-002-B Wall 2.04E-01 47.2 2.28E-01 52.8 RCA-TURB-003-B Wall 2.26E-01 50.0 2.26E-01 50.0 RCA-TURB-004-B Wall 6.34E-01 35.1 1.17E+00 64.9 RCA-TURB-005-B Wall 8.55E-01 32.4 1.78E+00 67.6 RCA-TURB-006-B Wall 1.48E+00 64.9 7.99E-01 35.1 RCA-TURB-007-B Wall 3.95E+00 69.5 1.73E+00 30.5 RCA-TURB-008-B Wall 1.96E-01 50.5 1.92E-01 49.5 RCA-TURB-009-B Wall 2.16E-01 48.0 2.34E-01 52.0 RCA-TURB-010-B Wall 9.19E-01 80.1 2.29E-01 19.9 RCA-TURB-011-B Wall 2.06E-01 77.2 6.07E-02 22.8 RCA-TURB-012-B Wall 1.10E+03 95.7 4.89E+01 4.3 RCA-TURB-013-B Wall 4.03E+00 62.4 2.43E+00 37.6 RCA-TURB-014-B Wall 2.07E+01 72.9 7.70E+00 27.1 RCA-TURB-015-B Wall 7.12E-02 55.2 5.79E-02 44.8 RCA-TURB-016-B Wall 8.14E-02 25.5 2.38E-01 74.5 RCA-TURB-017-B Wall 1.79E+03 99.7 5.54E+00 0.3 RCA-TURB-018-B Wall 1.12E+03 99.7 3.44E+00 0.3 RCA-TURB-019-B Floor 1.70E-01 41.8 2.37E-01 58.2 RCA-TURB-020-B Floor 6.13E-02 36.0 1.09E-01 64.0 RCA-TURB-021-B Floor 1.85E-01 44.4 2.32E-01 55.6 RCA-TURB-022-B Floor 5.29E-01 58.3 3.78E-01 41.7 RCA-TURB-023-B Floor 5.15E+02 93.9 3.37E+01 6.1 RCA-TURB-024-B Floor 8.31E+02 92.4 6.81E+01 7.6 RCA-TURB-025-B Floor 9.07E+01 98.2 1.63E+00 1.8 RCA-TURB-026-B Floor 2.49E+02 77.7 7.13E+01 22.3 RCA-TURB-027-B Floor 3.63E+02 91.4 3.40E+01 8.6 RCA-TURB-028-B Floor 4.38E+00 84.2 8.24E-01 15.8 RCA-TURB-029-B Floor 2.90E+00 79.3 7.56E-01 20.7 RCA-TURB-030-B Floor 1.07E+02 87.3 1.56E+01 12.7 RCA-TURB-031-B Floor 2.56E+02 94.1 1.61E+01 5.9 RCA-TURB-032-B Floor 2.68E+02 95.3 1.33E+01 4.7 RCA-TURB-033-B Floor 1.48E+02 96.0 6.13E+00 4.0 RCA-TURB-034-B Floor 1.04E+02 92.0 9.08E+00 8.0 RCA-TURB-035-B Floor 1.50E+03 96.6 5.29E+01 3.4 RCA-TURB-036-B Floor 3.19E+03 94.3 1.93E+02 5.7 RCA-TURB-037-B Floor 5.13E+02 69.5 2.25E+02 30.5 RCA-TURB-038-B Floor 8.37E+01 86.2 1.34E+01 13.8 RCA-TURB-039-B Floor 2.26E+02 89.2 2.73E+01 10.8 RCA-TURB-040-B Floor 1.14E+02 88.8 1.44E+01 11.2 RCA-TURB-041-B Floor 2.19E+02 88.0 2.99E+01 12.0

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Location Cs-137 % Cs-137 Co-60 % Co-60 (pCi/g) (pCi/g)

RCA-TURB-042-B Floor 1.89E+03 92.7 1.48E+02 7.3 RCA-TURB-043-B Floor 3.87E+02 97.5 9.72E+00 2.5 RCA-TURB-044-B Floor 1.74E+03 85.2 3.03E+02 14.8 RCA-TURB-045-B Floor 6.66E+02 66.9 3.29E+02 33.1 Mean 75 Mean 25

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= 10.7525 4.70E + 04+ 0.1.35E + 04

= 2.90E + 04 dpm100cm2 Where:

0.75 is the mean fraction for Cs-137 4.70E+04 is the DCGL for Cs-1 37 0.25 is the mean fraction for Co-60 1.35E+04 is the DCGL for Co-60 4.3 New Radwaste Building Tunnel Concrete Results OCNGS New Radwaste Tunnel Concrete was characterized from 08/29/2023 to 08/30/2023. Nine samples were taken from biased wall and floor surface area locations. Cs-137 and Co-60 were the only plant-related gamma radionuclides positively identified, and Table 4 provides the analytical results of this effort.

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Table 4: OCNGS New Radwaste Tunnel Concrete Assessment Location Cs-137 % Cs-137 Co-60 % Co-60 (pCi/g) (pCi/g)

RCA-NRWT-001-B Wall 1.81E+01 57.3 1.35E+01 42.7 RCA-NRWT-002-B Floor 6.58E+02 97.4 1.79E+01 2.6 RCA-NRWT-003-B Floor 3.27E+01 94.7 1.83E+00 5.3 RCA-NRWT-004-B Floor 1.54E+01 81.8 3.43E+00 18.2 RCA-NRWT-005-B Floor 4.82E+01 82.3 1.04E+01 17.7 RCA-NRWT-006-B Floor 8.74E+01 87.4 1.26E+01 12.6 RCA-NRWT-007-B Floor 5.42E+03 59.2 3.73E+03 40.8 RCA-NRWT-008-B Wall 3.07E-01 35.7 5.52E-01 64.3 RCA-NRWT-009-B Wall 4.43E+00 32.6 9.14E+00 67.4 Mean 70 Mean 30

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= 10.7030 4.70E + 04+ 0.1.35E + 04

= 2.69E + 04 dpm100cm2

Where:

0.70 is the mean fraction for Cs-137 4.70E+04 is the DCGL for Cs-1 37 0.30 is the mean fraction for Co-60 1.35E+04 is the DCGL for Co-60

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4.4 Gross Alpha DCGL In the absence of positive data on the fractional composition of alpha emitters at OCNGS, the conservatively determined DCGL of 1650 dpm/100 cm2 for Am-241, the most limiting prevalent alpha emitter at OCNGS, will be utilized when conducting alpha surveys.

4.5 Modification of DCGLGA values Decommissioning activities and unforeseen events have the potential to cause changes in the fractions of radionuclides, contributing to gross activity.

Radionuclide variability can occur due to changes in operational procedures, environmental conditions, and contamination pathways. For example, areas with higher water ingress or flooding might display a higher proportion of Cs-137 than Co-60, influencing the site-specific DCGLGA. Continued characterization will address these differences. Additionally, radionuclide ratios may vary throughout the major structures. As a result, areassessment of the DCGLGA values is necessary during the data quality objective (DQO) phase of survey plan development, particularly when ongoing characterization efforts reveal a change in radionuclide fractions or when a particular area within a structure exhibits a different radionuclide variability (i.e., different radionuclide ratios).

5 Assumptions

  • Radionuclide fractions are derived from biased sample locations and may not represent the total area uniformly.
  • The DCGLGA values calculated are based on gamma emitters Cs-137 and Co-60, with negligible contributions from beta and alpha emitters.
  • The variability in sample results could affect the accuracy of the DCGLGA, necessitating re-evaluation during future characterization phases.
  • The conservative assumptions may result in overestimating contamination in some areas.

6 Conclusions

Each structure-specific DCGL GA will be assessed during the DQO phase of the Survey Unit FSS plan development. This reassessment is critical to account for changes in radionuclide fractions or operational conditions that may impact radiological contamination levels.

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The unity rule will be used to assess survey results showing both beta/gamma and alpha activity. This ensures that all relevant radionuclides are accurately accounted for when determining compliance with release criteria.

The initial DCGL for administrative purposes or the Operational DCGL (DCGL op) may be set to 60% of the Listed DCGL for compliance with the State of New Jersey guidance, further reinforcing conservative safety measures.

Continuous monitoring and evaluation of the DCGLGA values, especially during decommissioning activities, is essential to ensure ongoing compliance with regulatory standards and to adapt to any unforeseen changes in contamination or environmental factors.

7 References

NRC. (2022), Characterization, Survey and Determination of Radiological Criteria Volume 2 Revision 2.

Westinghouse, (2024), Oyster Creek Nuclear Generating Station Below Grade Structures Radiological Characterization Report Revision 0.

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