ML25345A132
| ML25345A132 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/11/2025 |
| From: | Holtec Decommissioning International, Radiation Safety & Control Services |
| To: | Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML25345A092 | List:
|
| References | |
| HDI 25-039 24-062, Rev 1 | |
| Download: ML25345A132 (0) | |
Text
1 l P a g e RSCS Technical Support Document No.24-062 Rev 1 Method for Determining Nuclide Fractions and Gross Activity DCGLs Prepared by:
Marty Erickson Reviewed by:
Christopher Messier Reviewed by:
Bill Parish Approved by:
Eric Darois Radiation Safety & Control Services, Inc 93 Ledge Road Seabrook, NH 03874 1-800-525-8339 www.radsafety.com April 02, 2025 RSCS Radiation Safety & Control Services
2 l P a g e 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......................................................................................... 10 4.5 Modification of DCGLGA values....................................................................... 10 5
Assumptions........................................................................................................... 10 6
Conclusions............................................................................................................ 11 7
References............................................................................................................. 11 List of Tables Table 1: OCNGS Building Surface DCGLs..................................................................... 4 Table 2: OCNGS Reactor Building Concrete Assessment.............................................. 5 Table 3: OCNGS Turbine Building Concrete Assessment.............................................. 7 Table 4: OCNGS New Radwaste Tunnel Concrete Assessment.................................... 9
3 l P a g e 1 Executive Summary This document outlines the methodology for calculating Gross Activity Derived Concentration Guideline Levels (DCGLGA) 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, with a focus 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 DCGLGA 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, along with its conservative assumptions, ensures that the decommissioning of OCNGS adheres to regulatory standards, providing a clear path for safely releasing the site for future use.
4 l P a g e 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 DCGLGA should be set to a value considering the plant-related radionuclide fractions. Calculating the DCGLGA 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 Nuclide 25 mrem/y DCGLa (dpm/100cm2) 15 mrem/y DCGLb (dpm/100cm2)
Nuclide 25 rem/y DCGLa (dpm/100cm2) 15 mrem/y DCGLb (dpm/100cm2)
Am-241 5.81E+02 3.49E+02 Nb-94 1.98E+04 1.19E+04 C-14 3.58E+06 2.15E+06 Ni-63 4.78E+06 2.87E+06 Cm-243 8.60E+02 5.16E+02 Np-237 4.70E+02 2.82E+02 Cm-244 1.08E+03 6.48E+02 Pu-238 6.70E+02 4.02E+02 Co-60 1.36E+04 8.16E+03 Pu-239 5.92E+02 3.55E+02 Cs-137 4.27E+04 2.56E+04 Pu-240 5.93E+02 3.56E+02 Eu-152 2.85E+04 1.71E+04 Pu-241 2.02E+04 1.21E+04 Eu-154 2.62E+04 1.57E+04 Sb-125 7.86E+04 4.72E+04 Fe-55 1.18E+07 7.08E+06 Sr-90 1.83E+04 1.10E+04 H-3 1.32E+09 7.92E+08 Tc-99 1.78E+06 1.07E+06 Mn-54 5.56E+04 3.34E+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:
5 l P a g e
- 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 1 +
2 2 +
Where:
f1, f2,, fn are the mean fractions of the radionuclides DCGL1, DCGL2,, DCGLn 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 samples 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.
Table 2: OCNGS Reactor Building Concrete Assessment Location Cs-137 (pCi/g)
% Co-60 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
6 l P a g e Location Cs-137 (pCi/g)
% Co-60 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 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 The variability in the Cs-137 and Co-60 percentages is likely due to the nature of biased sample collection, which targets areas expected to have higher activity concentrations. The high
7 l P a g e percentage of Cs-137 in certain samples may reflect localized contamination due to operational history in those areas.
Fifty-seven samples sent off-site for HTD analysis found that the pure beta-emitting radionuclides (i.e., Sr-90, Ni-63, H-3, and C-14) were positively identified and contributed less than 0.03 to the fraction of radionuclides. Using only the gamma emitters Cs-137 and Co-60, the resultant DCGLGA was approximately 500 dpm/100 cm² lower and, therefore, more conservative.
Based on the analysis of the data presented in Table 2, the value of DCGLGA for walls and floor in the OCNGS Reactor Building can be determined as follows:
Equation 2: Calculation of the Reactor Building DCGLGA
=
1 0.85 4.27E + 04 +
0.15 1.36E + 04
= 3.23E + 04 dpm 100cm2 Where:
0.85 is the mean fraction for Cs-137 4.27E+04 is the DCGL for Cs-1 37 0.15 is the mean fraction for Co-60 1.36E+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.
Table 3: OCNGS Turbine Building Concrete Assessment Location Cs-137 (pCi/g)
% Co-60 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
8 l P a g e Location Cs-137 (pCi/g)
% Co-60 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 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 The variability in the Cs-137 and Co-60 percentages is likely due to the nature of biased sample collection, which targets areas expected to have higher activity concentrations. The high percentage of Cs-137 in certain samples may reflect localized contamination due to operational history in those areas.
Fifty-four samples sent off-site for HTD analysis found that the pure beta-emitting radionuclides (i.e., Sr-90 and C-14) were positively identified, contributing less than 0.006 to the fraction of
9 l P a g e radionuclides. By using only the gamma emitters Cs-137 and Co-60, the resultant DCGLGA was lower and, therefore, more conservative.
Based on the analysis of the data presented in Table 3, the value of DCGLGA for walls and floor in the OCNGS Turbine Building can be determined as follows:
Equation 3: Calculation of the Turbine Building DCGLGA
=
1 0.75 4.27E + 04 +
0.25 1.36E + 04
= 2.78E + 04 dpm 100cm2 Where:
0.75 is the mean fraction for Cs-137 4.27E+04 is the DCGL for Cs-1 37 0.25 is the mean fraction for Co-60 1.36E+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.
Table 4: OCNGS New Radwaste Tunnel Concrete Assessment Location Cs-137 (pCi/g)
% Co-60 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 The variability in the Cs-137 and Co-60 percentages is likely due to the nature of biased sample collection, which targets areas expected to have higher activity concentrations. The high percentage of Cs-137 in certain samples may reflect localized contamination due to operational history in those areas.
10 l P a g e Ten samples sent off-site for HTD analysis found that pure beta-emitting radionuclides (i.e., Sr-90 and C-14) were positively identified in one sample, contributing 0.006 to the fraction of radionuclides. Using only the Cs-137 and Co-60 gamma emitters, the resultant DCGLGA was lower and, therefore, more conservative.
Based on the analysis of the data presented in Table 4, the value of DCGLGA for walls and floor in the OCNGS New Radwaste Tunnel can be determined as follows:
Equation 4: Calculation of the New Radwaste Tunnel DCGLGA
=
1 0.70 4.27E + 04 +
0.30 1.36E + 04
= 2.60E + 04 dpm 100cm2 Where:
0.70 is the mean fraction for Cs-137 4.27E+04 is the DCGL for Cs-1 37 0.30 is the mean fraction for Co-60 1.36E+04 is the DCGL for Co-60 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 581 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, a reassessment 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.
11 l P a g e 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 DCGLGA 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.
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 (DCGLop) 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, are 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.