ML23240A537

From kanterella
Jump to navigation Jump to search
Confirmatory Survey Activities Summary and Results for the Unit 2 and 3 Intake Structures at the San Onofre Nuclear Generating Station San Clemente CA
ML23240A537
Person / Time
Site: San Onofre  Southern California Edison icon.png
Issue date: 08/18/2023
From: Bailey E
Oak Ridge Institute for Science & Education
To: Amy Snyder
Reactor Decommissioning Branch
References
5365-SR-01-0, RFTA 22-004
Download: ML23240A537 (1)
v • d • e


Text

August 18, 2023 Ms. Amy Snyder Senior Project Manager U.S. Nuclear Regulatory Commission 11545 Rockville Pike Mail Stop A10M Rockville, MD 20852

SUBJECT:

CONFIRMATORY SURVEY ACTIVITIES

SUMMARY

AND RESULTS FOR THE UNIT 2 AND 3 INTAKE STRUCTURES AT THE SAN ONOFRE NUCLEAR GENERATING STATION, SAN CLEMENTE, CALIFORNIA DOCKET NUMBER 50-361 AND 50-362; RFTA NO.22-004; DCN: 5365-SR-01-0

Dear Ms. Snyder:

The Oak Ridge Institute for Science and Education (ORISE) is pleased to provide the enclosed final report detailing the confirmatory survey activities summary and results performed for the Unit 2 and 3 Intake Structures at the San Onofre Nuclear Generating Station in San Clemente, California. Comments on the draft report have been addressed in this final version.

Please feel free to contact me at Erika.Bailey@orau.org if you have any comments or concerns.

Sincerely, Erika N. Bailey Survey Projects Manager ORISE ENB:enb electronic distribution: T. Barvitskie, NRC L. Gersey, NRC G. Chapman, NRC C. Dennes, NRC S. Anderson, NRC K. Engel, ORISE D. Hagemeyer, ORISE File/5365 100 ORAU Way

  • Oak Ridge
  • TN 37830
  • orise.orau.gov

CONFIRMATORY SURVEY ACTIVITIES

SUMMARY

AND RESULTS FOR THE UNIT 2 AND 3 INTAKE STRUCTURES AT THE SAN ONOFRE NUCLEAR GENERATING STATION, SAN CLEMENTE, CALIFORNIA K. M. Engel ORISE FINAL REPORT Prepared for the U.S. Nuclear Regulatory Commission AUGUST 2023 Further dissemination authorized to NRC only; other requests shall be approved by the originating facility or higher NRC programmatic authority.

ORAU provides innovative scientific and technical solutions to advance research and education, protect public health and the environment and strengthen national security. Through specialized teams of experts, unique laboratory capabilities and access to a consortium of more than 100 major Ph.D.-granting institutions, ORAU works with federal, state, local and commercial customers to advance national priorities and serve the public interest. A 501(c) (3) nonprofit corporation and federal contractor, ORAU manages the Oak Ridge Institute for Science and Education (ORISE) for the U.S. Department of Energy (DOE). Learn more about ORAU at www.orau.org.

NOTICES The opinions expressed herein do not necessarily reflect the opinions of the sponsoring institutions of Oak Ridge Associated Universities.

This report was prepared as an account of work sponsored by the United States Government.

Neither the United States Government nor the U.S. Department of Energy, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe on privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, mark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement or recommendation, or favor by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.

CONFIRMATORY SURVEY ACTIVITIES

SUMMARY

AND RESULTS FOR THE UNIT 2 AND 3 INTAKE STRUCTURES AT THE SAN ONOFRE NUCLEAR GENERATING STATION, SAN CLEMENTE, CALIFORNIA FINAL REPORT Prepared by K. M. Engel ORISE AUGUST 2023 Prepared for the U.S. Nuclear Regulatory Commission This document was prepared for the U.S. Nuclear Regulatory Commission (NRC) by the Oak Ridge Institute for Science and Education (ORISE) through Interagency Agreement 31310021S0033 between the NRC and the U.S. Department of Energy (DOE). ORISE is managed by Oak Ridge Associated Universities (ORAU) under DOE contract number DE-SC0014664.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report 5365-SR-01-0

CONFIRMATORY SURVEY ACTIVITIES

SUMMARY

AND RESULTS FOR THE UNIT 2 AND 3 INTAKE STRUCTURES AT THE SAN ONOFRE NUCLEAR GENERATING STATION, SAN CLEMENTE, CALIFORNIA Prepared by: Date: 8/18/2023 K. M. Engel, CHP, Health Physicist ORISE Reviewed by: Date: 8/18/2023 N. A. Altic, CHP, Health Physicist ORISE 8/18/2023 Reviewed by: Date:

E. N. Bailey, Survey Projects Manager ORISE Reviewed by: Date: 8/18/2023 P. H. Benton, Quality Manager ORISE Reviewed and approved for release by: Date: 8/18/2023 Derek Hagemeyer, Director, IEAV ORISE FINAL REPORT AUGUST 2023 SONGS Unit 2 and 3 ISA Confirmatory Survey Report i 5365-SR-01-0

CONTENTS FIGURES .......................................................................................................................................................... iii TABLES ............................................................................................................................................................. iii ACRONYMS .................................................................................................................................................... iv

1. INTRODUCTION....................................................................................................................................... 6
2. SITE DESCRIPTION ................................................................................................................................. 7
3. DATA QUALITY OBJECTIVES ............................................................................................................. 9 3.1 State the Problem ............................................................................................................................10 3.2 Identify the Decision/Objective ...................................................................................................10 3.3 Identify Inputs to the Decision/Objective..................................................................................11 3.3.1 Radionuclides of Concern and Release Guidelines............................................................12 3.4 Define the Study Boundaries .........................................................................................................14 3.5 Develop a Decision Rule................................................................................................................14 3.5.1 PSQ1: Comparison Against Applicable Limits ..................................................................15 3.5.2 PSQ2: SU Classification .........................................................................................................15 3.6 Specify Limits on Decision Errors ...............................................................................................16 3.7 Optimize the Design for Obtaining Data....................................................................................17
4. PROCEDURES ..........................................................................................................................................17 4.1 Reference System ............................................................................................................................17 4.2 Surface Scans....................................................................................................................................17 4.3 Surface Activity Measurements .....................................................................................................18 4.4 Surface Removable Activity Measurements ................................................................................18 4.5 Volumetric Sampling ......................................................................................................................18
5. SAMPLE ANALYSIS AND DATA INTERPRETATION ...............................................................19
6. FINDINGS AND RESULTS ...................................................................................................................20 6.1 Surface Scans....................................................................................................................................20 6.2 Radionuclide Concentrations in Volumetric Samples ...............................................................21
7.

SUMMARY

AND CONCLUSIONS ......................................................................................................24

8. REFERENCES ...........................................................................................................................................25 APPENDIX A: FIGURES APPENDIX B: DATA TABLES APPENDIX C: MAJOR INSTRUMENTATION APPENDIX D: SURVEY AND ANALYTICAL PROCEDURES SONGS Unit 2 and 3 ISA Confirmatory Survey Report ii 5365-SR-01-0

FIGURES Figure 2.1. SONGS Site Layout....................................................................................................................... 8 Figure 2.2. Unit 2 and 3 Intake Structure Area ............................................................................................. 9 TABLES Table 3.1. Independent Confirmatory Survey Decision Process ..............................................................11 Table 3.2. Unit 2 and 3 Intake Structure Concrete EDCGLs ...................................................................13 Table 6.1. SONGS Intakes Unit 2 and 3 Scan Data Summary .................................................................21 Table 6.2. SONGS Intakes Unit 2 and 3 Summary of Concrete Sample Data ......................................23 SONGS Unit 2 and 3 ISA Confirmatory Survey Report iii 5365-SR-01-0

ACRONYMS Am-241 americium-241 C-14 carbon-14 cm centimeter(s)

Cm-243 curium-243 Cm-244 curium-244 Co-60 cobalt-60 cpm counts per minute Cs-134 cesium-134 Cs-137 cesium-137 DCGL derived concentration guideline level dpm/100 cm2 disintegrations per minute per 100 square centimeters DQO data quality objective EDCGL estimated derived concentration guideline level Eu-152 europium-152 Eu-154 europium-154 Fe-55 iron-55 FSS final status survey H-3 tritium H0 null hypothesis HA alternative hypothesis IL investigation level ISA Intake Structure Area ISFSI independent spent fuel storage installation LTP License Termination Plan m2 square meter(s)

MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDC minimum detectable concentration mg/cm2 milligram per square centimeter MLLW mean lower low water mrem/yr millirem per year NaI[Tl] thallium-doped sodium iodide Nb-94 niobium-94 Ni-59 nickel-59 Ni-63 nickel-63 NIST National Institute of Standards and Technology Np-237 neptunium-237 NRC U.S. Nuclear Regulatory Commission ORAU Oak Ridge Associated Universities ORISE Oak Ridge Institute for Science and Education pCi/g picocurie per gram Pm-147 promethium-147 PSQ principal study question Pu-238 plutonium-238 Pu-239 plutonium-239 Pu-240 plutonium-240 SONGS Unit 2 and 3 ISA Confirmatory Survey Report iv 5365-SR-01-0

Pu-241 plutonium-241 Pu-242 plutonium-242 RESL Radiological and Environmental Sciences Laboratory ROC radionuclide of concern Sb-125 antimony-125 SCE Southern California Edison SOF sum-of-fractions SONGS San Onofre Nuclear Generating Station Sr-90 strontium-90 SU survey unit Tc-99 technetium-99 TEDE total effective dose equivalent UGLLT ultima gold LLT SONGS Unit 2 and 3 ISA Confirmatory Survey Report v 5365-SR-01-0

CONFIRMATORY SURVEY ACTIVITIES

SUMMARY

AND RESULTS FOR THE UNIT 2 AND 3 INTAKE STRUCTURES AT THE SAN ONOFRE NUCLEAR GENERATING STATION SAN CLEMENTE, CALIFORNIA

1. INTRODUCTION The San Onofre Nuclear Generating Station (SONGS) consisted of three pressurized water reactors with a total rated capacity of 2,664 net megawatts electrical. Unit 1, rated at 410 net megawatts electrical, was supplied by Westinghouse Electric Company and operated from January 1968 until permanent shutdown in November 1992. All above-ground structures were dismantled for Unit 1 and all fuel was transferred to the independent spent fuel storage installation (ISFSI) in 2004. By 2006, all remaining monitored effluent pathways from Unit 1 were permanently removed or routed to Unit 2 for discharge to the outfall. The remaining portions of Unit 1 are owned by Southern California Edison (SCE) and San Diego Gas and Electric. Units 2 and 3, rated at 1,127 net megawatts electrical each, were supplied by Combustion Engineering, Inc. and are owned by SCE, San Diego Gas and Electric, and the City of Riverside. The units began commercial operation in August 1983 and April 1984, respectively. On June 12, 2013, SCE notified the U.S. Nuclear Regulatory Commission (NRC) that Units 2 and 3 had permanently ceased operations. By August 2020, all spent fuel had been transferred from the spent fuel pools of Units 2 and 3 to the ISFSI (SCE 2021).

The site utilized a once-through circulating water system that withdrew cooling water from and discharged to the Pacific Ocean. Units 2 and 3 each have their own intake structure, co-located as mirror images in an area known as the Intake Structure Area (ISA). Each intake structure housed the circulating water system, associated pumps, plus the saltwater cooling pumps. The intake and discharge tunnels are attached to the intake structure (Areva 2015). Low concentrations of radionuclides of concern (ROCs) have been positively detected in samples taken from the intakes attributed to planned and monitored discharges from Unit 1 as documented in the Historical Site Assessment and 2015 Site Characterization Report, after an event and from effluent sampling in 1997, and a personnel contamination event in 2020. Following final status surveys (FSS), the intake structures for Units 2 and 3 are scheduled to be backfilled with flowable, low-density grout to an elevation of seven-foot mean lower low water (MLLW) to support remediation and demolition of other site buildings. All sediment was removed from the intakes and the remaining intake structure SONGS Unit 2 and 3 ISA Confirmatory Survey Report 6 5365-SR-01-0

basements have been surveyed and sampled by the site prior to backfill using methods that meet the requirements of a FSS (SDS 2022a).

NRC requested that Oak Ridge Institute for Science and Education (ORISE) perform confirmatory survey activities of the Units 2 and 3 ISA. NRC will use the confirmatory survey data for their evaluation of the FSS data relative to the project end-point criteria.

2. SITE DESCRIPTION SONGS is located in San Clemente, California, in northern San Diego County next to San Onofre State Beach. The site is approximately 102 kilometers (64 miles) south of Los Angeles and 82 kilometers (51 miles) north-northwest of San Diego. The SONGS site lies entirely within the boundaries of the Marine Corp Base Camp Pendleton under a grant of easement between SCE and the United States Government. The site is bounded on the west by the Pacific Ocean, on the east by Interstate Highway 5, and on both the north and south along the coastline by San Onofre State Beach (Enercon 2014). Figure 2.1 depicts the layout of the SONGS site.

The ISA is an approximate 3,400 square meters (m2) area located on the west side of the Unit 2 and 3 protected area between the Turbine Buildings. Essentially all the intake/outfall system is below grade (i.e., below the 30-foot elevation). The intake and discharge tunnels are attached to the intake structure (Areva 2015). Figure 2.2 depicts the Unit 2 and 3 ISA. To dewater the ISA for FSS, a grout bag was permanently placed on the Unit 2 return side to isolate the Pacific Ocean from the Unit 2 Intake Structure while accommodating continued use of the saltwater dilution pumps. Gates 3 and 5 were installed on the intake side to isolate the remaining Unit 2 area. Unit 3 was isolated from the Pacific Ocean by installing Stop Gate 1 (return side) and Gates 3 and 5 on the intake side. Following FSS, a flowable backfill will be dispensed from the foot up to the 7-foot elevation in both the Unit 2 and 3 Intake Structures. It is intended that the grout bag and the backfilled area will remain in place after license termination (SDS 2022a).

The foot and foot floor surfaces and conjoined wall surfaces up to 1-meter for both units were classified by the site as Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) Class 2 survey units (SUs). The area between the outfall weir and Gate 1 up to the ceiling were Class 2 SUs. The remaining areas were classified as Class 3. The foot and foot elevations have a floor surface area (excluding the grout bag footprint) of 1,358 m2 in the Unit 2 SONGS Unit 2 and 3 ISA Confirmatory Survey Report 7 5365-SR-01-0

Intake Structure and 1,395 m2 in the Unit 3 Intake Structure; therefore, a minimum of two Class 2 SUs and one Class 3 SU were required for each intake structure (SDS 2022a).

Figure 2.1. SONGS Site Layout (Enercon 2014)

SONGS Unit 2 and 3 ISA Confirmatory Survey Report 8 5365-SR-01-0

Figure 2.2. Unit 2 and 3 Intake Structure Area (Areva 2015)

3. DATA QUALITY OBJECTIVES The data quality objectives (DQOs) described herein are consistent with the Guidance on Systematic Planning Using the Data Quality Objectives Process (EPA 2006) and provided a formalized method for planning radiation surveys, improving survey efficiency and effectiveness, and ensuring that the type, quality, and quantity of data collected were adequate for the intended decision applications. The seven steps in the DQO process were as follows:
1. State the problem
2. Identify the decision/objective
3. Identify inputs to the decision/objective
4. Define the study boundaries
5. Develop a decision rule
6. Specify limits on decision errors
7. Optimize the design for obtaining data SONGS Unit 2 and 3 ISA Confirmatory Survey Report 9 5365-SR-01-0

Confirmatory survey DQOs were originally presented in ORISE 2023 and are represented here for completeness.

3.1 STATE THE PROBLEM The first step in the DQO process defined the problem that necessitated the study, identified the planning team, and examined the project budget and schedule. The planning team, project budget, and schedule are presented in ORISE 2023 and are not discussed here. Objectives of the confirmatory survey were to provide NRC with independent radiological data to assist NRC in evaluating the FSS results. Therefore, the problem statement is as follows:

Confirmatory survey activities are necessary to generate independent radiological data to assist NRC with their assessment of the FSS design, implementation, and results for demonstrating compliance with the release criteria for the Unit 2 and 3 ISA.

3.2 IDENTIFY THE DECISION/OBJECTIVE The second step in the DQO process identified the principal study questions (PSQs) and alternative actions, developed decision statements, and organized multiple decisions, as appropriate. This second step is done by specifying alternative actions that could result from a Yes response to the PSQs and combining the PSQs and alternative actions into decision statements. Given that the problem statement introduced in Section 3.1 is fairly broad, multiple PSQs arose. PSQs, alternative actions, and combined decision statements are presented in Table 3.1. As originally planned, the first PSQ related to the degree to which the FSS data and confirmatory data agreed, while the second PSQ focused on confirming the appropriateness of the SU classification. Due to the complexity of the intake structures SUs, ORISE did not generate the necessary data for comparison against the FSS results. Instead, ORISE only collected judgmental confirmatory measurement data to assess whether residual radioactivity concentrations were below applicable limits. The collected data were sufficient to evaluate PSQ2, therefore subsequent DQOs addressing PSQ2 were unchanged from the project-specific plan.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report 10 5365-SR-01-0

Table 3.1. Independent Confirmatory Survey Decision Process Principal Study Questions Alternative Actions Yes:

Compile confirmatory data and report results to NRC for their decision making. Provide independent interpretation that confirmatory field surveys did not identify anomalous areas of residual radioactivity, quantitative field and laboratory data satisfied NRC-approved decommissioning PSQ1: Do the confirmatory survey results criteria, and/or that statistical sample population agree with the FSS data for structural examination/assessment conditions were met.

surfaces and volumetric concentrations and are residual radioactivity concentrations No:

below applicable limits? Compile confirmatory data and report results to NRC for their decision making. Provide independent interpretation of confirmatory survey results, identifying any anomalous field or laboratory data and/or when statistical sample population examination/assessment conditions were not satisfied for NRCs determination of the adequacy of the FSS data.

Yes:

Confirmatory results support the classification of the FSS SU. Compile confirmatory survey data and present results to NRC for their decision making.

PSQ2: Do the confirmatory results support the MARSSIM classification of the FSS SU?

No:

Confirmatory results do not support the classification of the FSS SU. Summarize the discrepancies and provide technical comments to NRC for their decision making.

Decision Statements Confirmatory survey results for volumetric concentrations, surface activity levels, and removable contamination [did/did not] identify anomalous results or other conditions that would preclude the data from demonstrating compliance with the release criteria, as applicable.

Confirmatory survey results [do/do not] support the sites MARSSIM classification of the FSS SUs.

3.3 IDENTIFY INPUTS TO THE DECISION/OBJECTIVE The third step in the DQO process identified both the information needed and the sources of this information, determined the basis for action levels, and identified sampling and analytical methods that will meet data requirements. For this effort, information inputs include the following:

  • FSS plans and data;
  • Derived concentration guideline levels (DCGLs), further discussed in subsection 3.3.1; SONGS Unit 2 and 3 ISA Confirmatory Survey Report 11 5365-SR-01-0
  • ORISE confirmatory survey results for surface radiation scans and surface activity levels;
  • ORISE volumetric sample and removable activity results.

3.3.1 Radionuclides of Concern and Release Guidelines To determine the acceptability of the survey results prior to backfill of the intake structures for Unit 2 and 3, DCGLs that were considered applicable to intake structure concrete were estimated. The DCGLs were calculated to represent a residual radioactive contamination level, above background, which would result in a total effective dose equivalent (TEDE) of 15 millirem per year (mrem/yr) to an average member of the critical group. The 15 mrem/yr limit is an administrative unrestricted use criterion applied at SONGS. It is noted that the current SONGS strategy is to ultimately apply the 25 mrem/yr criterion from Title 10 Code of Federal Regulations 20 Subpart E in the License Termination Plan (LTP) which will be the basis for NRC review and approval. To avoid confusion, the term EDCGL is used to refer to the estimated DCGLs and to indicate they are not final. The site considers the EDCGLs conservative values used for survey planning and assessment with confidence that the DCGLs ultimately provided in the LTP will be higher. The potential presence of sediment with residual radioactivity was not considered in the EDCGLs for the Unit 2 and 3 Intake Structures since all sediment was removed. The EDCGL values for concrete are presented in Table 3.2 noting the calculations included an adjustment factor of 0.3 as an estimate of the a priori fraction that ultimately will be assigned in the LTP to calculate Operational DCGLs for basement concrete.

These EDCGLs represent 4.5 mrem/yr. The remaining 70% of the 15 mrem/yr is reserved as a margin to accommodate the dose contributions from media other than basement concrete; the final values assigned in the LTP may be different (SDS 2022a).

SONGS Unit 2 and 3 ISA Confirmatory Survey Report 12 5365-SR-01-0

Table 3.2. Unit 2 and 3 Intake Structure Concrete EDCGLs a Excavation All Concrete Below 7-Foot ROC Elevation Adjusted EDCGL (pCi/g)

Americium-241 (Am-241) 19.5 Carbon-14 (C-14) 50.0 Curium-243 (Cm-243) 49.5 Curium-244 (Cm-244) 65.9 Cobalt-60 (Co-60) 23.2 Cesium-134 (Cs-134) 27.2 Cesium-137 (Cs-137) 43.6 Europium-152 (Eu-152) 69.4 Europium-154 (Eu-154) 63.9 Iron-55 (Fe-55) 75,300 Tritium (H-3) 1,070 Niobium-94 (Nb-94) 21.7 Nickel-59 (Ni-59) 6,380 Nickel-63 (Ni-63) 2,330 Neptunium-237 (Np-237) 0.362 Promethium-147 (Pm-147) 82,500 Plutonium-238 (Pu-238) 13.7 Plutonium-239 (Pu-239) 12.4 Plutonium-240 (Pu-240) 12.4 Plutonium-241 (Pu-241) 611 Plutonium-242 (Pu-242) 13.0 Antimony-125 (Sb-125) 154 Strontium-90 (Sr-90) 4.50 Technetium-99 (Tc-99) 29.6 a Recreated from Table 4-1 in SDS 2022a ROC = radionuclide of concern EDCGL = estimated derived concentration guideline level pCi/g = picocuries per gram If the DCGLs were final, each individual DCGL would correspond to the TEDE criterion and the sum-of-fractions (SOF) approach would be used to evaluate the total dose from each sample to demonstrate compliance with the dose limit. However, with EDCGLs, it may be acceptable if a SOF value is greater than 1. The site noted that follow-up investigation actions would result if the EDCGLs or SOFs were exceeded (SDS 2022a). The SOF calculation is performed as follows:

SONGS Unit 2 and 3 ISA Confirmatory Survey Report 13 5365-SR-01-0

= =1 Eq. (3-1)

Where:

Cj is the concentration of ROC j DCGLj is the DCGL for ROC j 3.4 DEFINE THE STUDY BOUNDARIES The fourth step in the DQO process defined target populations and spatial boundaries, determined the timeframe for collecting data and making decisions, addressed practical constraints, and determined the smallest subpopulations, area, volume, and time for which separate decisions must be made.

NRC identified the Unit 2 and 3 ISA for confirmatory survey activities. FSS activities had already been completed in the Unit 3 Intake Structure before the confirmatory site visit, and site staff were still collecting data in the Unit 2 Intake Structure during confirmatory activities. Confirmatory activities were performed in both Unit 2 and 3 during the dates of January 23-26, 2023. The temporal boundaries included the overall project schedule. The specific timing of on-site confirmatory activities was dependent on the decommissioning contractors and the overall project schedule.

The physical boundary for the confirmatory activities was the intake structure concrete that extended from the 7-foot elevation down to the foot elevation in both units. This included SUs 1, 2, 3, 4, 5, and 6. The physical SU boundaries were dependent on safe access considerations. Based on previous site survey and sampling results, areas of interest included the outfall vent shaft area where former liquid radioactive waste effluent lines discharged into the outfall and the floor and lower walls of both units.

3.5 DEVELOP A DECISION RULE The fifth step in the DQO process specified appropriate population parameters (e.g., mean, median), developed action levels, confirmed detection limits are less than action levels, and developed an ifthen decision rule statement.

Two PSQs were introduced in Table 3.1; therefore, two decision rules arose. As previously SONGS Unit 2 and 3 ISA Confirmatory Survey Report 14 5365-SR-01-0

mentioned in Section 3.2, PSQ1 was revised based on on-site survey conditions. Therefore, the associated decision rule was modified and is discussed further in Section 3.5.1.

3.5.1 PSQ1: Comparison Against Applicable Limits Any positive detection required direct comparison against the applicable EDCGL, thus a formal statistical comparison to the original dataset was unnecessary. The confirmatory survey focused on high density scanning of select areas to identify locations that could exceed the applicable release criterion and/or analytical minimum detectable concentrations (MDCs). Qualitative parameters of interest included gamma and alpha-plus-beta scan data. The quantitative parameters of interest were sample concentrations and the associated MDCs. The decision rule addressing PSQ1 is stated as:

If each individual sample result is below the applicable DCGL, then conclude that residual radioactivity concentrations are acceptable for demonstrating compliance with the release criterion; otherwise, perform further evaluation(s) and provide technical comments/recommendations to NRC for their evaluation and decision making.

3.5.2 PSQ2: SU Classification PSQ2, identified in Table 3.1, sought to confirm whether the SUs were correctly classified and/or whether a particular SU should have been reclassified as a result of the FSS and was not.

Reclassification relates primarily to Class 2 and Class 3 SUs, as well as non-impacted areas, as a Class 1 SU will not receive a higher classification. Site areas were categorized as either radiologically impacted or non-impacted. Non-impacted areas have no reasonable potential for residual contamination. Impacted SUs were classified based on contamination potential, as either Class 1, 2, or 3 in accordance with MARSSIM (NRC 2000). A description of each impacted SU classification is as follows:

Class 1: Areas that had a significant potential for radioactive contamination (based on site operating history) or known contamination that exceeds the DCGLs.

Class 2: Areas, often contiguous to Class 1 areas, which had a potential for radioactive contamination but at levels less than the DCGLs.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report 15 5365-SR-01-0

Class 3: Remaining impacted areas that were not expected to contain residual contamination or were expected to contain levels of residual contamination at a small fraction of the DCGLs.

For the intake structures, on-site confirmatory investigation levels (ILs) included any areas with a reproducible indication above background identified by the surveyor in real time during scanning.

Post-survey ILs would have included any exceedances of the EDCGLs for Class 2 SU samples or levels at a small fraction of the EDCGLs for Class 3 SU samples. However, as previously noted, it may be acceptable to exceed an EDCGL or a SOF value greater than 1 and still meet the 25 mrem/yr criterion.

The decision rule addressing PSQ2 is stated as:

If surface activity levels or volumetric concentrations indicate that a Class 2 or Class 3 SU should be reclassified to a higher classification, summarize confirmatory data for NRCs evaluation.

3.6 SPECIFY LIMITS ON DECISION ERRORS The sixth step in the DQO process examined the consequences of making an incorrect decision and established bounds on decision errors. Decision errors were controlled during the survey design, on-site field investigations, and during laboratory analysis. These controls were related to the scanning, measurement, and analytical MDCs.

Surface scan MDCs were calculated to ensure a true positive detection proportion of 0.95 and a false positive proportion of no more than 0.25 for beta and gamma radiation. The surface scans combined with liberal pausing were effective for locating potential residual contamination hot spots.

As a second order of control, analytical procedures and count times ensured that detection limits were a fraction of the release criteria and are discussed in Appendix D. Field scanning MDCs were minimized by following the procedures referenced in Sections 4 and 5, respectively. Scan and static MDC information for hand-held detectors are presented in Appendix D. Additionally, analytical MDCs of 10% of the each DCGL were requested.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report 16 5365-SR-01-0

3.7 OPTIMIZE THE DESIGN FOR OBTAINING DATA The seventh step in the DQO process was used to review DQO outputs, develop data collection design alternatives, formulate mathematical expressions for each design, select the sample size to satisfy DQOs, decide on the most resource-effective design of agreed alternatives, and document requisite details. Specific survey procedures are presented in Section 4. The confirmatory survey followed a graded approach, such that survey efforts were concentrated in the areas that had the most likely potential for contamination within the SUs.

4. PROCEDURES The ORISE survey team performed visual inspections, measurements, and sampling activities within the intake structures during the dates of January 23-26, 2023. Survey activities were conducted in accordance with the project-specific IV survey plan, the Oak Ridge Associated Universities (ORAU)

Radiological and Environmental Survey Procedures Manual, and the ORAU Environmental Services and Radiation Training Quality Program Manual (ORISE 2023, ORAU 2016, ORAU 2022). Appendices C and D provide additional information regarding survey instrumentation and related processes discussed within this section.

4.1 REFERENCE SYSTEM ORISE referenced confirmatory measurement/sampling locations to prominent intake structure features. Measurement and sampling locations were documented on detailed survey maps.

4.2 SURFACE SCANS Surface scans were performed with Ludlum model 44-10, 5.1-centimeter by 5.1-centimeter (2-inch by 2-inch) thallium-doped sodium iodide (NaI[Tl]), hereafter referred to as NaI, scintillation detectors for gamma radiation. Ludlum model 43-37 floor monitor and 43-68 hand-held gas-flow proportional detectors with 0.8 milligram per square centimeter (mg/cm2)-thick Mylar windows were used for scanning in alpha-plus-beta mode. Alpha-only scans were not performed due to time constraints and because the alpha-plus-beta scans did not identify areas of contamination with exception of the elevated radiation levels in the vent shaft area near the former liquid radioactive waste effluent lines. All detector types were coupled to Ludlum model 2221 ratemeter-scalers with audible indicators and coupled to hand-held data-loggers to electronically record detector response.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report 17 5365-SR-01-0

Locations of elevated response that were audibly distinguishable from localized background levels, suggesting the presence of residual contamination, were marked for further investigation.

Confirmatory scan density was medium- to high-density (50% to 100% scan coverage) on floors, lower walls, and on the upper walls in only the vent shafts.

4.3 SURFACE ACTIVITY MEASUREMENTS Alpha-plus-beta surface activity measurements were performed with Ludlum model 43-68 hand-held gas-flow proportional detectors with 0.8 mg/cm2-thick Mylar windows. The detectors were paired with Ludlum model 2221 ratemeter-scalers with audible output. Surface activity measurements were only collected at the judgmental sample locations prior to the collection of three volumetric concrete samples. Material-specific background surface activity measurements were not collected with the hand-held detectors as there were no DCGLs developed/presented in units of disintegrations per minute per 100 square centimeters (dpm/100 cm2) to compare to surface activity results. The ORISE results presented are gross surface activity values.

4.4 SURFACE REMOVABLE ACTIVITY MEASUREMENTS After consultation with NRC, smear samples for determining removable alpha/beta activity levels were not collected from direct measurement locations. Volumetric samples were collected from the direct measurement locations for direct comparison to the EDCGLs presented in Table 3.2.

4.5 VOLUMETRIC SAMPLING Prior to the on-site confirmatory activities, NRC requested that 21 volumetric samples collected by site staff prior to installation of the grout bag in the Unit 2 Intake Structure be sent to ORISE for confirmatory analysis. These samples were re-labeled with an ORISE sample ID for independent laboratory analysis. Refer to sample IDs 5365M0001 through 5365M0021 in Table B.1 which also includes the sites sample IDs. The samples were shipped from GEL Laboratories, LLC in Charleston, South Carolina to the ORISE facility in Oak Ridge, TN.

During on-site confirmatory activities, three judgmental concrete core samples were collected by site staff under supervision of ORISE staff. These three samples, sample IDs 5365M0022 through 5365M0024 in Table B.1, were collected at locations exhibiting the highest elevated direct radiation levels identified by ORISE during scans. Samples 5365M0022 and 5365M0023 were collected from SONGS Unit 2 and 3 ISA Confirmatory Survey Report 18 5365-SR-01-0

the upper walls in the Unit 2 Intake Structure in the vent shaft area that once contained a liquid effluent discharge line. Sample 5365M0024 was collected from the upper wall in the Unit 3 Intake Structure, also in the vent shaft area that once contained a liquid effluent discharge line. All sampling equipment was cleaned/rinsed in the field after the collection of each sample to prevent cross-contamination. The cores were packaged by ORISE field staff by location indicating the top side of each core. The top 0.5 inch of each core sample was analyzed.

Following on-site activities, six additional FSS samples were shipped from GEL Laboratories to the ORISE facility for confirmatory analysis per NRCs request. The FSS samples were re-labeled with ORISE sample IDs 5365M0025 through 5365M0030; Table B.1 also includes the sites sample IDs.

The first two judgmental samples received were collected from the same location in the Unit 3 Intake Structure, near the top of the vent shaft; sample 5365M0025 represents the top 0.5-inch segment of the core and sample 5365M0026 is the 0.5-inch - 1.0-inch segment of the core. Sample IDs 5365M0027 through 5365M0030 were random sample locations collected from the Unit 2 circulating water system conduits that ran from the 7-foot elevation of the Unit 2 Intake Structure to the Unit 2 Condenser inside the Turbine Building and back to the Intake Structure at the foot elevation. The homogenized crushed core media that was received represents the top 0.5 inch of each core location.

5. SAMPLE ANALYSIS AND DATA INTERPRETATION Data collected on-site were transferred to the ORISE facility for analysis and interpretation. All samples were transferred under chain-of-custody to the Radiological and Environmental Sciences Laboratory (RESL) in Idaho Falls, Idaho. Sample analyses were performed in accordance with the laboratorys applicable procedures or other special instructions from ORISE. Volumetric samples were homogenized and a portion of each sample was set aside for other potential future analyses before being dried and analyzed by gamma spectrometry for gamma-emitting fission and activation products. Select samples were analyzed for additional ROCs listed in Table 3.2 following discussions and approval from NRC staff. C-14 was not reported as RESL does not currently have an approved procedure for performing C-14 analysis in solid samples. RESL does not typically report Pu-241 via gamma spectrometry when mainly statistical zeros are reported for Pu-238, and Pu-239/240 as it should not be present. Due to their similar alpha energies, curium results are reported as Cm-243/244 and plutonium results are reported as Pu-239/240. Additionally, after the review of the SONGS Unit 2 and 3 ISA Confirmatory Survey Report 19 5365-SR-01-0

gamma spectrometry data, NRC did not opt for further alpha spectrometry or other hard to detect analysis in every sample. Analytical results were reported in units of picocuries per gram (pCi/g).

Surface scan results are presented as gross counts per minute (cpm). Surface activity measurements are reported as gross dpm/100 cm2.

6. FINDINGS AND RESULTS The results of the confirmatory survey activities are discussed in the following subsections.

6.1 SURFACE SCANS Table 6.1 summarizes the scan data collected in the Unit 2 and 3 Intake Structures on the floors, lower walls, and upper walls of the vent shafts. Figures A.1 through A.6 in Appendix A present the sites maps of the survey units for completeness. Elevated count rates relative to background were noted in SUs 5 and 6, which are the upper walls in the vent shafts in Units 2 and 3, respectively.

These elevated areas were marked for further investigation and volumetric samples were collected.

Sustained elevated count rates were not observed in the other SUs.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report 20 5365-SR-01-0

Table 6.1. SONGS Intakes Unit 2 and 3 Scan Data Summary Detector Intake Response SU Class Area Radiation Type (cpm)

Unit Min Max

-26' Elev Floors Gamma 9,000 16,000 SU 1 and 2 2 -26' Elev Floors + 540 1,700

-26' Elev Lower Walls + 130 680

-13' and -7' Elev Floors Gamma --a --a SU 2 2 -13' and -7' Elev Floors +b 81 700 Unit 2

-13' and -7' Elev Lower Walls + 33 570 South and West Upper Walls Gamma 11,000 27,000 in vent shaft SU 5 3 South and West Upper Walls

+ 130 1,280 in vent shaft

-26' Elev Floors Gamma 10,800 15,700 SU 3 and 4 2 -26' Elev Floors + 400 1,680

-26' Elev Lower Walls + 110 680

-13' and -7' Elev Floors Gamma 3,800 16,000 SU 4 2 -13' and -7' Elev Floors + 300 1,400

-13' and -7' Elev Lower Walls + 10 740 Unit 3 North and West Upper Walls Gamma 10,000 17,000 in vent shaft North and West Upper Walls SU 6 3 + 110 1,300 in vent shaft Above the flowable fill line, Gamma 13,000 30,000 concrete will be removed a Surveyor could not recall exact scan range, but no elevated locations of concern were noted during scans.

b All + floor scans were performed with Ludlum model 43-37 floor monitor except for SU 2 -13' and -7' Elev Floors which were performed with Ludlum model 43-68 hand-held detector.

6.2 RADIONUCLIDE CONCENTRATIONS IN VOLUMETRIC SAMPLES NRC requested that a total of 27 volumetric concrete samples collected by the site be sent to ORISE for confirmatory analysis, including sample IDs 5365M0001 through 5365M0021 collected prior to installation of the grout bag in the Unit 2 Intake Structure, sample IDs 5365M0025 and 5365M0026 collected in the Unit 3 Intake Structure near the top of the vent shaft SU, concrete that will remain in place, and sample IDs 5365M0027 through 5365M0030 collected from the Unit 2 circulating water system conduits. Additionally, three volumetric concrete samples were collected during the confirmatory site visit from judgmental locations identified by ORISE during scans. Two of the samples, 5365M0022 and 5365M0023, were collected from Unit 2 SU 5 from the upper south and west walls in the vent shaft area. Sample 5365M0024 was collected from Unit 3 SU 6 from the SONGS Unit 2 and 3 ISA Confirmatory Survey Report 21 5365-SR-01-0

upper west wall in the vent shaft area. Table B.1 in Appendix B presents the analytical results for each individual concrete sample and the calculated SOFs. Table B.2 presents the gross surface activity levels collected prior to the collection of sample IDs samples, 5365M0022 through 5365M0024. The surface activity values presented are assuming all activity is from Cs-137 because the analytical results for each sample had statistically positive results for Cs-137.

Table 6.2 provides a summary of the ROC concentrations for all samples analyzed. All samples had a SOF valuebased on the EDCGLs presented in Table 3.2less than unity, which means that individual ROC concentrations were less than their respective EDCGL. Based on current industry guidance, such as in MARSSIM Section N.4, all reported concentrations greater than zero, that are not considered statistically positive at the 95% confidence interval, were included in the SOF calculations. Additionally, negative values were treated as zeros. Only the ROCs that were analyzed were included in the ORISE SOF calculations. For the curium and plutonium results, the most conservative DCGLs were used in determining the SOF. That is, for Cm-243/244 the Cm-243 DCGL of 49.5 pCi/g was used and for Pu-239/240 the Pu-239 DCGL of 12.4 pCi/g was used noting the DCGL for Pu-240 is also 12.4 pCi/g. For reasons stated in section 5, C-14 and Pu-241 are not accounted for in the SOF calculations. RESL does not currently have an approved procedure for C-14 analysis in solid samples and Pu-241 is not typically reported via gamma spectrometry when statistical zeros are reported for Pu-238 and Pu-239/240, which is the case for most of the confirmatory samples. With a DCGL of 611 pCi/g, it is not expected for Pu-241 to significantly contribute to the SOF.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report 22 5365-SR-01-0

Table 6.2. SONGS Intakes Unit 2 and 3 Summary of Concrete Sample Data No. of Samples No. of Statistically Concentrations (pCi/g)

Parameter Analyzed Positive Results Min Max Am-241 30/30 2 -0.04 0.02 C-14a 0/30 0 -- --

Cm-243/244b 9/30 0 -0.005 0.002 Co-60 30/30 2 -0.6 0.7 Cs-137 30/30 5 -0.4 2.82 Eu-152 30/30 1 -1.2 1.0 Eu-154 30/30 0 -1.2 2.3 Fe-55 12/30 1 -0.6 1.0 H-3 9/30 0 -0.04 90 Nb-94 30/30 0 -0.3 0.3 Ni-59 9/30 0 -0.60 0.17 Ni-63 12/30 1 -0.60 1.0 Np-237 9/30 1 -0.60 0.04 Pm-147c 4/30 0 -0.0002 0.009 Pu-238d 12/30 4 -0.003 0.0287 Pu-239/240d,e 12/30 4 -0.009 0.0135 Pu-241f 0/30 0 -- --

Pu-242 5/30 0 -0.001 0.0041 Sb-125 30/30 0 -0.5 3 Sr-90 12/30 0 -0.26 0.7 Tc-99 9/30 0 -0.40 0.11 SOF -- -- 0.00005 0.23 aThe selected laboratory does not currently have an approved C-14 procedure for solid samples.

bDue to similar alpha energies, the curium results reported as Cm-243/244. The most conservative DCGL was used in the SOF calculation (i.e., Cm-243).

cNRC requested Pm-147 be reported for samples 5367M0027 - 5367M0030 only. Results not provided for the remainder of samples.

dSee shaded footnote on Table B.1.

eDue to similar alpha energies, Pu results reported as Pu-239/240. The most conservative DCGL was used in the SOF calculation (i.e., Pu-239) noting the DCGL is the same for Pu-239 and Pu-240.

fAnalysis not performed after review of the Pu-238 and Pu-239/240 results.

The confirmatory concentrations for the ROCs that were analyzed have concentrations well below the EDCGLs. The sites SOF range for the 21 samples, 5365M0001 through 5365M00021, collected prior to installation of the grout bag in the Unit 2 Intake Structure, is 0.035 to 0.138 (SDS 2022b) versus the SOF range of 0.00005 to 0.11 for the confirmatory analysis of these samples (the confirmatory analysis did not include all the ROCs). The SOF for the four conduit samples analyzed, 5367M0027 through 5365M0030, ranged from 0.13 to 0.23 for the confirmatory analysis.

For the FSS sample core collected in the Unit 3 Intake Structure near the top of the vent shaft, SONGS Unit 2 and 3 ISA Confirmatory Survey Report 23 5365-SR-01-0

represented by 5365M0025 for the top 0.5 inch of the core, the sites Cs-137 reported concentration was 3.25 pCi/g versus the confirmatory result of 2.82 pCi/g both of which are less than 10% of the EDGCL. For the three judgmental sample locations identified by ORISE, the SOF range was 0.01 to 0.05 and the maximum Cs-137 concentration was 0.549 pCi/g. Additionally, all confirmatory sample concentrations for reported ROCs were less than 10% of the respective EDCGLs; therefore, the confirmatory sample concentrations did not indicate that a Class 2 or Class 3 SU should be reclassified to a higher classification.

7.

SUMMARY

AND CONCLUSIONS The ORISE survey team performed independent visual inspections, measurements, and sampling activities within the Unit 2 and Unit 3 Intake Structures including SUs 1, 2, 3, 4, 5, and 6 during the period of January 23-26, 2023. The confirmatory activities consisted of gamma scans, alpha-plus-beta scans, alpha-plus-beta direct measurements, and volumetric sampling in select areas. Based on the results of the scans performed on select areas of the floor, lower walls, and upper walls, three judgmental volumetric samples were collected from locations as described previously. ORISE encountered no evidence of discrete radioactive particles during confirmatory activities. Additionally, NRC requested that 27 volumetric concrete samples collected by site staff be sent to ORISE for confirmatory analysis. In summary, the confirmatory concentrations for the ROCs that were analyzed have concentrations well below the EDCGLs. The sites maximum reported Cs-137 concentration was 3.25 pCi/g versus the confirmatory result of 2.82 pCi/g for the same sample.

Additionally, all confirmatory sample concentrations for reported ROCs were less than 10% of the respective EDCGLs. All samples had a SOF value less than unity (i.e., <1) with a maximum SOF of 0.23.

Structural surface scans identified little elevated activity. When volumetric concentrations are at or below the MDC, statical comparisons between datasets would not be beneficial. The confirmatory results were all less than 10% of the EDCGLs which support the current SU classification.

ORISE did not identify any anomalous issues from the areas investigated that would preclude the sites data from demonstrating compliance with the EDCGLs. Furthermore, the confirmatory survey data supports the SU classification.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report 24 5365-SR-01-0

8. REFERENCES Areva 2015. San Onofre Nuclear Generating Station Historical Site Assessment Report. 39-9229156-002.

Prepared by Bartlett Nuclear Energy, Plymouth, Massachusetts, and URS Corporation. La Jolla, California. February.

Enercon 2014. SONGS Units 2 and 3 Environmental Impact Evaluation. Revision 0. Enercon Services, Inc. January 29.

EPA 2006. Guidance on Systematic Planning Using the Data Quality Objectives Process. EPA QA/G-4.

U.S. Environmental Protection Agency. Washington, D.C. February.

NRC 2000. Multi-Agency Radiation Site Survey and Investigation Manual (MARSSIM), Revision 1.

DOE/EH-0624, Rev. 1. U.S. Department of Energy. Washington, D.C. August 2000.

NRC 2020. Minimum Detectable Concentrations with Typical Radiation Survey for Instruments for Various Contaminants and Field Conditions, Revision 1. U.S. Nuclear Regulatory Commission.

Washington, D.C. August.

ORAU 2016. ORAU Radiological and Environmental Survey Procedures Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. November 10.

ORAU 2020a. ORAU Health and Safety Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. October 29.

ORAU 2020b. ORAU Radiation Protection Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. November 17.

ORAU 2022. ORAU Environmental Services and Radiation Training Quality Program Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. September 8.

ORISE 2023. Project-Specific Plan for Confirmatory Survey Activities for the Unit 2 and 3 Intake Structures at the San Onofre Nuclear Generating Station, San Clemente, California. Oak Ridge Institute for Science and Education. Oak Ridge, Tennessee. January 17.

SCE 2021. San Onofre Nuclear Generating Station 2021 Annual Radiological Environmental Operating Report. Southern California Edison.

SDS 2022a. SONGS Technical Support Document Unit 2 and 3 Intake Structure Final Status Survey Supplement San Onofre Nuclear Generating Station. SCS-RP1-TSD-22-09, Rev. 4. SDS and AECOM Energy Solutions Joint Venture. October.

SDS 2022b. SONGS FSS Memorandum U2 Intake Structure Grout Bag Region Data Quality Analysis From Dale Randall, LT/FSS Technical Manager. SDS and AECOM Energy Solutions Joint Venture. June 15.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report 25 5365-SR-01-0

APPENDIX A: FIGURES SONGS Unit 2 and 3 ISA Confirmatory Survey Report 5365-SR-01-0

Figure A.1. Unit 2 SU 1, shaded in red SONGS Unit 2 and 3 ISA Confirmatory Survey Report A-1 5365-SR-01-0

Figure A.2. Unit 2 SU 2, shaded in red SONGS Unit 2 and 3 ISA Confirmatory Survey Report A-2 5365-SR-01-0

Figure A.3. Unit 2 SU 5, shaded in red SONGS Unit 2 and 3 ISA Confirmatory Survey Report A-3 5365-SR-01-0

Figure A.4. Unit 3 SU 3, shaded in red SONGS Unit 2 and 3 ISA Confirmatory Survey Report A-4 5365-SR-01-0

Figure A.5. Unit 3 SU 4, shaded in red SONGS Unit 2 and 3 ISA Confirmatory Survey Report A-5 5365-SR-01-0

Figure A.6. Unit 3 SU 6, shaded in red SONGS Unit 2 and 3 ISA Confirmatory Survey Report A-6 5365-SR-01-0

APPENDIX B: DATA TABLES SONGS Unit 2 and 3 ISA Confirmatory Survey Report 5365-SR-01-0

Table B.1. Volumetric Sample Concentrations ORISE Sample ID 5365M0001 5365M0002 5365M0003 5365M0004 5365M0005 5365M0006 5365M0007 Site Sample ID C9002-101C1-SCV1-004 C9002-101C1-SCV1-005 C9002-101C1-SCV1-006 C9002-101C1-SCV1-007 C9002-101C1-SCV1-008 C9002-101C1-SCV1-009 C9002-101C1-SCV1-013 Conc.a TPUb Conc. TPU Conc. TPU Conc. TPU Conc. TPU Conc. TPU Conc. TPU Am-241 (pCi/g) 0.007 0.022 0.008 0.012 0.02 0.08 -0.009 0.014 0.015 0.024 0.003 0.008 0.02 0.12 C-14 (pCi/g)c -- -- -- -- -- -- -- -- -- -- -- -- -- --

Cm-243/244 (pCi/g)d -- -- -- -- -- -- -- -- -- -- -- -- -- --

Co-60 (pCi/g) 0.002 0.008 0.017 0.024 -0.003 0.008 0.015 0.02 0.007 0.008 -0.6 0.8 -0.05 0.38 Cs-137 (pCi/g) 0.0002 0.0006 0.010 0.022 0.004 0.026 0.008 0.052 0.006 0.006 -0.3 0.4 0.3 1 Eu-152 (pCi/g) -0.017 0.022 0.5 5.0 0.002 0.012 -0.004 0.012 0.008 0.056 -0.4 2.0 0.2 2.0 Eu-154 (pCi/g) -0.002 0.012 0.6 2.2 0.2 0.6 0.019 0.034 -0.021 0.028 0.2 1.2 1.0 2.0 Fe-55 (pCi/g) 0.5 2.2 -- -- -- -- -- -- 0.00 2.20 -- -- -- --

H-3 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Nb-94 (pCi/g) -0.004 0.006 0.003 0.022 -0.001 0.012 0.004 0.014 0.005 0.008 -0.3 0.8 -0.3 0.6 Ni-59 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Ni-63 (pCi/g) 0.7 1.2 1.0 1.2 Np-237 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Pm-147 (pCi/g)e -- -- -- -- -- -- -- -- -- -- -- -- -- --

Pu-238 (pCi/g) -0.003 0.012 -- -- -- -- -- -- 0.003 0.012 -- -- -- --

Pu-239/240 (pCi/g)f 0.011 0.18 -- -- -- -- -- -- 0.006 0.014 -- -- -- --

Pu-241 (pCi/g) g -- -- -- -- -- -- -- -- -- -- -- -- -- --

Pu-242 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Sb-125 (pCi/g) 0.016 0.022 0.016 0.028 -0.006 0.014 -0.01 0.024 -0.01 0.10 0.6 2.6 -0.23 0.26 Sr-90 (pCi/g) 0.1 0.6 -- -- -- -- -- -- 0.2 0.6 -- -- -- --

Tc-99 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

SOF 0.02 -- 0.02 -- 0.004 -- 0.001 -- 0.05 -- 0.01 -- 0.03 --

aResults that are considered statistically positive are bolded.

bUncertainties are based on total propagated uncertainties at the 95% confidence level; 2-sigma uncertainty is presented.

cThe selected laboratory does not currently have an approved C-14 procedure for solid samples.

dDue to similar alpha energies, the curium results reported as Cm-243/244. The most conservative DCGL was used in SOF calculation (i.e., Cm-243).

eOnly reported for sample IDs 5365M0027 - 5365M0030; potential interferences at very low concentrations via gamma spec.

fDue to similar alpha energies, Pu results reported as Pu-239/240. The most conservative DCGL was used in SOF calculation (i.e., Pu-239) noting the DCGL is the same for Pu-239 and Pu-240.

gAnalysis not performed after review of the Pu-238 and Pu-239/240 results.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report B-1 5365-SR-01-0

Table B.1. Volumetric Sample Concentrations (Continued)

ORISE Sample ID 5365M0008 5365M0009 5365M0010 5365M0011 5365M0012 5365M0013 5365M0014 Site Sample ID C9002-101C1-SCV1-014 C9002-101C1-QCV1-014 C9002-101C1-SCV1-015 C9002-101C1-SCV1-016A C9002-101C1-SCV1-017A C9002-101C1-SCV1-018A C9002-101C1-SCV1-001 Conc.a TPUb Conc. TPU Conc. TPU Conc. TPU Conc. TPU Conc. TPU Conc. TPU Am-241 (pCi/g) 0.016 0.018 0.02 0.06 0.01 0.04 0.006 0.008 0.002 0.010 -0.012 0.024 0.003 0.012 C-14 (pCi/g) c -- -- -- -- -- -- -- -- -- -- -- -- -- --

Cm-243/244 (pCi/g)d -- -- -- -- -- -- -- -- -- -- -- -- -- --

Co-60 (pCi/g) -0.3 0.8 0.7 0.8 0.005 0.008 0.2 0.8 0.0005 0.0012 0.08 0.58 0.2 0.8 Cs-137 (pCi/g) 0.2 0.4 0.3 1.2 0.008 0.010 0.3 0.6 0.001 0.008 -0.4 0.6 0.02 0.08 Eu-152 (pCi/g) -1.2 1.8 -0.1 1.0 0.007 0.024 1.0 3.2 0.05 0.04 -0.1 1.0 -1.1 2.8 Eu-154 (pCi/g) -0.03 0.18 2.3 2.8 -0.011 0.034 1.7 2.4 -0.07 0.08 -1.2 2.4 1.4 2.4 Fe-55 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

H-3 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Nb-94 (pCi/g) -0.1 0.6 0.3 1.8 0.006 0.054 -0.2 0.8 -0.004 0.010 0.1 0.8 -0.09 0.66 Ni-59 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Ni-63 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Np-237 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Pm-147 (pCi/g)e -- -- -- -- -- -- -- -- -- -- -- -- -- --

Pu-238 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Pu-239/240 (pCi/g) f -- -- -- -- -- -- -- -- -- -- -- -- -- --

Pu-241 (pCi/g)g -- -- -- -- -- -- -- -- -- -- -- -- -- --

Pu-242 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Sb-125 (pCi/g) -0.5 3.2 3.0 3.2 0.6 1.2 0.05 0.18 -0.09 0.12 0.3 1.8 0.03 0.30 Sr-90 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Tc-99 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

SOF 0.01 -- 0.11 -- 0.01 -- 0.06 -- 0.001 -- 0.01 -- 0.03 --

aResults that are considered statistically positive are bolded.

bUncertainties are based on total propagated uncertainties at the 95% confidence level; 2-sigma uncertainty is presented.

cThe selected laboratory does not currently have an approved C-14 procedure for solid samples.

dDue to similar alpha energies, the curium results reported as Cm-243/244. The most conservative DCGL was used in SOF calculation (i.e., Cm-243).

eOnly reported for sample IDs 5365M0027 - 5365M0030; potential interferences at very low concentrations via gamma spec.

fDue to similar alpha energies, Pu results reported as Pu-239/240. The most conservative DCGL was used in SOF calculation (i.e., Pu-239) noting the DCGL is the same for Pu-239 and Pu-240.

gAnalysis not performed after review of the Pu-238 and Pu-239/240 results.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report B-2 5365-SR-01-0

Table B.1. Volumetric Sample Concentrations (Continued)

ORISE Sample ID 5365M0015 5365M0016 5365M0017 5365M0018 5365M0019 5365M0020 5365M0021 Site Sample ID C9002-101C1-SCV1-002 C9002-101C1-SCV1-003 C9002-101C1-JCV1-001 C9002-101C1-JCV1-002 C9002-101C1-SCV1-010 C9002-101C1-SCV1-011 C9002-101C1-SCV1-012 Conc.a TPUb Conc. TPU Conc. TPU Conc. TPU Conc. TPU Conc. TPU Conc. TPU Am-241 (pCi/g) 0.010 0.036 0.01 0.12 -0.04 0.10 0.007 0.022 0.008 0.018 0.02 0.10 -0.004 0.012 C-14 (pCi/g)c -- -- -- -- -- -- -- -- -- -- -- -- -- --

Cm-243/244 (pCi/g) d -- -- -- -- -- -- -- -- -- -- -- -- -- --

Co-60 (pCi/g) 0.012 0.012 -0.005 0.012 -0.006 0.014 0.002 0.010 -0.003 0.010 0.0008 0.0056 -0.007 0.012 Cs-137 (pCi/g) -0.003 0.018 -0.004 0.008 -0.004 0.008 -0.009 0.008 0.057 0.010 -0.012 0.012 0.001 0.010 Eu-152 (pCi/g) -0.0009 0.0088 0.01 0.04 -0.010 0.034 -0.002 0.022 0.01 0.06 -0.017 0.024 -0.010 0.026 Eu-154 (pCi/g) 0.006 0.042 0.004 0.028 -0.023 0.030 -0.018 0.032 0.007 0.046 0.016 0.036 -0.002 0.016 Fe-55 (pCi/g) -- -- -- -- -- -- -0.6 2.2 -- -- -- -- -- --

H-3 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Nb-94 (pCi/g) -0.002 0.016 -0.002 0.008 0.001 0.010 0.004 0.008 0.006 0.012 -0.003 0.008 0.007 0.012 Ni-59 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Ni-63 (pCi/g) -- -- -- -- -- -- 0.5 1.2 -- -- -- -- -- --

Np-237 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Pm-147 (pCi/g) e -- -- -- -- -- -- -- -- -- -- -- -- -- --

Pu-238 (pCi/g) -- -- -- -- -- -- 0.003 0.012 -- -- -- -- -- --

Pu-239/240 (pCi/g)f -- -- -- -- -- -- 0.003 0.016 -- -- -- -- -- --

Pu-241 (pCi/g) g -- -- -- -- -- -- -- -- -- -- -- -- -- --

Pu-242 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

Sb-125 (pCi/g) -0.01 0.04 0.005 0.032 -0.007 0.028 -0.006 0.026 0.012 0.026 0.023 0.032 0.003 0.020 Sr-90 (pCi/g) -- -- -- -- -- -- 0.2 0.6 -- -- -- -- -- --

Tc-99 (pCi/g) -- -- -- -- -- -- -- -- -- -- -- -- -- --

SOF 0.001 -- 0.001 -- 0.00005 -- 0.05 -- 0.002 -- 0.001 -- 0.0004 --

aResults that are considered statistically positive are bolded.

bUncertainties are based on total propagated uncertainties at the 95% confidence level; 2-sigma uncertainty is presented.

cThe selected laboratory does not currently have an approved C-14 procedure for solid samples.

dDue to similar alpha energies, the curium results reported as Cm-243/244. The most conservative DCGL was used in SOF calculation (i.e., Cm-243).

eOnly reported for sample IDs 5365M0027 - 5365M0030; potential interferences at very low concentrations via gamma spec.

fDue to similar alpha energies, Pu results reported as Pu-239/240. The most conservative DCGL was used in SOF calculation (i.e., Pu-239) noting the DCGL is the same for Pu-239 and Pu-240.

gAnalysis not performed after review of the Pu-238 and Pu-239/240 results.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report B-3 5365-SR-01-0

Table B.1. Volumetric Sample Concentrations (Continued)

ORISE Sample ID 5365M0022 5365M0023 5365M0024 5365M0025 5365M0026 C9002-203F1-CV033-J Site Sample ID n/a n/a n/a C9002-203F1-CV033-J (2nd segment)

Conc.a TPUb Conc. TPU Conc. TPU Conc. TPU Conc. TPU Judgmental Samples Am-241 (pCi/g) -0.0015 0.0022 0.010 0.010 0.009 0.007 0.002 0.008 0.003 0.006 C-14 (pCi/g)c -- -- -- -- -- -- -- -- -- --

Cm-243/244 (pCi/g) d 0.0012 0.0044 0.001 0.005 -0.0002 0.0034 0.002 0.004 0.0007 0.0072 Co-60 (pCi/g) 0.02 0.04 0.008 0.012 0.006 0.008 0.004 0.004 0.007 0.016 Cs-137 (pCi/g) 0.009 0.006 0.549 0.028 0.393 0.022 2.82 0.14 -0.005 0.012 Eu-152 (pCi/g) -0.018 0.038 -0.009 0.020 0.008 0.012 0.005 0.044 0.017 0.078 Eu-154 (pCi/g) -0.018 0.030 -0.022 0.032 0.016 0.050 0.009 0.036 -0.001 0.006 Fe-55 (pCi/g) 0.07 0.38 -0.24 0.30 -0.3 0.4 -0.1 0.4 0.0 0.4 H-3 (pCi/g) -0.04 0.10 -0.01 0.10 -0.03 0.10 0.09 0.10 0.07 0.10 Nb-94 (pCi/g) -0.003 0.014 0.001 0.008 -0.005 0.010 -0.003 0.016 0.015 0.110 Ni-59 (pCi/g) -0.01 0.26 0.17 0.24 -0.13 0.30 0.08 0.32 0.09 0.34 Ni-63 (pCi/g) 0.18 0.30 0.4 0.4 0.17 0.34 0.38 0.38 0.36 0.38 Np-237 (pCi/g) 0.006 0.006 0.012 0.009 0.001 0.007 0.008 0.008 0.0014 0.0028 Pm-147 (pCi/g)e -- -- -- -- -- -- -- -- -- --

Pu-238 (pCi/g) 0.024 0.013 0.018 0.018 0.017 0.012 0.028 0.012 0.029 0.013 Pu-239/240 (pCi/g)f 0.007 0.005 0.014 0.010 -0.001 0.008 0.010 0.007 0.011 0.008 Pu-241 (pCi/g)g -- -- -- -- -- -- -- -- -- --

Pu-242 (pCi/g) 0.004 0.004 0.002 0.004 0.003 0.008 -0.001 0.004 0.0014 0.0028 Sb-125 (pCi/g) 0.011 0.032 -0.005 0.028 0.05 0.100 0.021 0.038 0.009 0.032 Sr-90 (pCi/g) -0.01 0.80 -0.1 0.8 -0.3 0.8 0.07 0.80 0.2 0.8 Tc-99 (pCi/g) -0.02 0.30 -0.19 0.30 -0.27 0.30 0.08 0.30 0.11 0.30 SOF 0.02 -- 0.05 -- 0.01 -- 0.11 -- 0.05 --

aResults that are considered statistically positive are bolded.

bUncertainties are based on total propagated uncertainties at the 95% confidence level; 2-sigma uncertainty is presented.

cThe selected laboratory does not currently have an approved C-14 procedure for solid samples.

dDue to similar alpha energies, the curium results reported as Cm-243/244. The most conservative DCGL was used in SOF calculation (i.e., Cm-243).

eOnly reported for sample IDs 5365M0027 - 5365M0030; potential interferences at very low concentrations via gamma spec.

fDue to similar alpha energies, Pu results reported as Pu-239/240. The most conservative DCGL was used in SOF calculation (i.e., Pu-239) noting the DCGL is the same for Pu-239 and Pu-240.

gAnalysis not performed after review of the Pu-238 and Pu-239/240 results.

Shaded results indicate that a separate run using a Pu-242 tracer and/or completion of a Pu-236 tracer check would be needed to confirm the absence or detection of plutonium at these lower levels.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report B-4 5365-SR-01-0

Table B.1. Volumetric Sample Concentrations (Continued)

ORISE Sample ID 5365M0027 5365M0028 5365M0029 5365M0030 Site Sample ID C8001-301F1-CV002-R-NRC C8001-301F1-CV004-NRC C8001-301F1-CV007-R-NRC C8001-301F1-CV010-R-NRC Conc.a TPUb Conc. TPU Conc. TPU Conc. TPU Am-241 (pCi/g) 0.008 0.012 -0.004 0.012 0.007 0.012 0.02 0.02 C-14 (pCi/g) c -- -- -- -- -- -- -- --

Cm-243/244 (pCi/g)d 0.0000 0.006 -0.002 0.004 -0.003 0.004 -0.005 0.008 Co-60 (pCi/g) 0.007 0.030 0.0014 0.0017 0.006 0.008 0.03 0.07 Cs-137 (pCi/g) -0.013 0.022 0.008 0.020 -0.010 0.022 -0.003 0.016 Eu-152 (pCi/g) -0.03 0.09 -0.2 0.6 -0.06 0.09 0.04 0.04 Eu-154 (pCi/g) -0.08 0.09 0.004 0.007 0.06 0.08 -0.008 0.045 Fe-55 (pCi/g) 1.0 1.0 0.4 0.8 -0.2 0.8 0.1 0.6 H-3 (pCi/g) 20 100 80 100 30 100 90 100 Nb-94 (pCi/g) 0.011 0.035 -0.022 0.026 -0.016 0.029 -0.003 0.017 Ni-59 (pCi/g) -0.1 0.6 -0.6 0.6 -0.2 0.6 -0.3 0.4 Ni-63 (pCi/g) -0.1 0.4 -0.6 0.6 -0.3 0.4 -0.3 0.4 Np-237 (pCi/g) 0.03 0.05 -0.6 1.0 0.04 0.04 0.022 0.025 Pm-147 (pCi/g)e -0.00017 0.00031 -0.0002 0.0005 -0.0001 0.0004 0.009 0.021 Pu-238 (pCi/g) 0.006 0.018 0.006 0.020 0.006 0.020 0.008 0.020 Pu-239/240 (pCi/g) f -0.005 0.018 -0.009 0.018 0.001 0.020 0.006 0.022 Pu-241 (pCi/g)g -- -- -- -- -- -- -- --

Pu-242 (pCi/g) -- -- -- -- -- -- -- --

Sb-125 (pCi/g) -0.04 0.06 0.02 0.06 0.007 0.060 -0.06 0.11 Sr-90 (pCi/g) 0.1 0.8 0.7 0.8 0.1 0.8 0.3 0.8 Tc-99 (pCi/g) -0.4 0.4 -0.2 0.4 -0.4 0.4 -0.4 0.4 SOF 0.13 -- 0.23 -- 0.16 -- 0.22 --

aResults that are considered statistically positive are bolded.

bUncertainties are based on total propagated uncertainties at the 95% confidence level; 2-sigma uncertainty is presented.

cThe selected laboratory does not currently have an approved C-14 procedure for solid samples.

dDue to similar alpha energies, the curium results reported as Cm-243/244. The most conservative DCGL was used in SOF calculation (i.e., Cm-243).

eOnly reported for sample IDs 5365M0027 - 5365M0030; potential interferences at very low concentrations via gamma spec.

fDue to similar alpha energies, Pu results reported as Pu-239/240. The most conservative DCGL was used in SOF calculation (i.e., Pu-239) noting the DCGL is the same for Pu-239 and Pu-240.

gAnalysis not performed after review of the Pu-238 and Pu-239/240 results.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report B-5 5365-SR-01-0

Table B.2. SONGS Intakes Unit 2 and 3 Surface Activity Levels at Concrete Sample Locations Total Beta Activity Sample ID Location (assuming all activity is from Cs-137) dpm/100 cm2 Unit 2, SU 5 5265M0022 2,600 South Upper Wall in Vent Shaft Unit 2, SU 5 5265M0023 4,100 West Upper Wall in Vent Shaft Unit 3, SU 6 5265M0024 3,900 West Upper Wall in Vent Shaft SONGS Unit 2 and 3 ISA Confirmatory Survey Report B-6 5365-SR-01-0

APPENDIX C: MAJOR INSTRUMENTATION SONGS Unit 2 and 3 ISA Confirmatory Survey Report 5365-SR-01-0

C.1. SCANNING AND MEASUREMENT INSTRUMENT/

DETECTOR COMBINATIONS The display of a specific product is not to be construed as an endorsement of the product or its manufacturer by the author or their employer.

C.1.1 GAMMA Ludlum NaI[Tl] Scintillation Detector Model 44-10, Crystal: 5.1 cm x 5.1 cm (Ludlum Measurements, Inc., Sweetwater, Texas)

Coupled to: Ludlum Ratemeter-scaler Model 2221 (Ludlum Measurements, Inc., Sweetwater, Texas)

Coupled to: Trimble Nomad (Trimble Navigation Limited, Sunnyvale, CA)

C.1.2 ALPHA-PLUS-BETA Ludlum Gas-flow Proportional Detector Model 43-68, 126 cm2 physical area, 0.8 mg/cm2 Mylar window (Ludlum Measurements, Inc., Sweetwater, Texas)

Coupled to: Ludlum Ratemeter-scaler Model 2221 (Ludlum Measurements, Inc., Sweetwater, Texas)

Coupled to: Trimble Nomad (Trimble Navigation Limited, Sunnyvale, CA)

Ludlum Gas-flow Proportional Detector Model 43-37, 584 cm2 physical area, 0.8 mg/cm2 Mylar window (Ludlum Measurements, Inc., Sweetwater, Texas)

Coupled to: Ludlum Ratemeter-scaler Model 2221 (Ludlum Measurements, Inc., Sweetwater, Texas)

Coupled to: Trimble Nomad (Trimble Navigation Limited, Sunnyvale, CA)

SONGS Unit 2 and 3 ISA Confirmatory Survey Report C-1 5365-SR-01-0

APPENDIX D: SURVEY AND ANALYTICAL PROCEDURES SONGS Unit 2 and 3 ISA Confirmatory Survey Report 5365-SR-01-0

D.1. PROJECT HEALTH AND SAFETY The Oak Ridge Institute for Science and Education (ORISE) performed all survey activities in accordance with the Oak Ridge Associated Universities (ORAU) Radiation Protection Manual, the ORAU Radiological and Environmental Survey Procedures Manual, and the ORAU Health and Safety Manual (ORAU 2020b, ORAU 2016, and ORAU 2020a). Prior to on-site activities, a Work-Specific Hazard Checklist was completed for the project and discussed with field personnel. The planned activities were thoroughly discussed with site personnel prior to implementation to identify hazards present.

Additionally, prior to performing work, a pre-job briefing and walk down of the survey areas were completed with field personnel to identify hazards present and discuss safety concerns. Should ORISE have identified a hazard not covered in ORAU 2016 or the projects Work-Specific Hazard Checklist for the planned survey and sampling procedures, work would not have been initiated or continued until the hazard was addressed by an appropriate job hazard analysis and hazard controls.

D.2. CALIBRATION AND QUALITY ASSURANCE Calibration of all field instrumentation was based on standards/sources traceable to National Institute of Standards and Technology (NIST).

Field survey activities were conducted in accordance with procedures from the following documents:

  • ORAU Radiological and Environmental Survey Procedures Manual (ORAU 2016)
  • ORAU Environmental Services and Radiation Training Quality Program Manual (ORAU 2022)

The procedures contained in these manuals were developed to meet the requirements of U.S. Department of Energy (DOE) Order 414.1D and U.S. Nuclear Regulatory Commissions (NRCs) Quality Assurance Manual for the Office of Nuclear Material Safety and Safeguards and contain measures to assess processes during their performance.

Quality control procedures include:

  • Daily instrument background and check-source measurements to confirm that equipment operation is within acceptable statistical fluctuations.
  • Training and certification of all individuals performing procedures.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report D-1 5365-SR-01-0

  • Periodic internal and external audits.

D.3. SURVEY PROCEDURES D.3.1 SURFACE SCANS Gamma scans were performed using Ludlum model 44-10 5.1-centimeter by 5.1-centimeter (2-inch by 2-inch) thallium-doped sodium iodide , referred to as NaI detectors. Alpha-plus-beta scans were performed using either a large-area (floor monitor) or hand-held gas-flow proportional detectors with a physical detector area of 584 square centimeters (cm2) or 126 cm2, respectively and 0.8 milligram per square centimeter (mg/cm2)-thick Mylar. Scans for elevated radiation were performed by passing the detector slowly over the surface. The distance between the detectors and surface was maintained at a minimum. Identification of elevated radiation levels that could exceed the localized background were determined based on an increase in the audible signal from the indicating instrument. The NaI gamma detectors were used solely as a qualitative means to identify elevated radiation levels in excess of background. However, for reference, NUREG-1507, Table 6-6, provides NaI scintillation detector scan minimum detectable concentrations (MDCs) for Common Radiological Contaminants (NRC 2020). For Cs-137, the scan MDC is 5.5 pCi/g and 2.8 pCi/g for Co-60. A specific scan MDC for the floor monitor was not determined as the instrument was used solely as a qualitative means to identify elevated radiation levels in excess of background and would have been quantitatively investigated using other hand-held instruments.

Surface scan MDCs for the hand-held gas-flow proportional detectors were estimated using the approach described in NUREG-1507 (NRC 2020). The scan MDC is a function of many variables, including a 1-second observation interval; a specified level of performance at the first scanning state of 95% true positive and 25% false positive rate, which yields a d' value of 2.32 (NUREG-1507, Table 6-1); and a surveyor efficiency of 0.5. For the structural surfaces, the conservative radionuclides of concern were assumed to be Co-60 for beta. As such the total efficiency for beta was 0.10. Only alpha-plus-beta scans were performed; therefore, the scan MDC was calculated using the following equation:

x x (/60) x (60/)

=

x x 100 2 SONGS Unit 2 and 3 ISA Confirmatory Survey Report D-2 5365-SR-01-0

Where:

d' = index of sensitivity

= background (cpm)

= observation interval (sec) p = surveyor efficiency

= total efficiency The scan MDC for surveys if assuming all activity is Co-60 is 3,400 disintegrations per minute (dpm)/100 cm2 based on an assumed background of 275 counts per minute (cpm) for the hand-held gas proportional detector. The scan MDC for surveys if assuming all activity is Cs-137 is 1,400 dpm/100 cm2 based on an assumed background of 275 cpm.

D.3.2 SURFACE ACTIVITY MEASUREMENTS Measurements of gross alpha and gross beta surface activity levels were performed using hand-held gas proportional detectors coupled to portable ratemeter-scalers. Count rates, which were integrated over 1 minute with the detector held in a static position, were converted to activity levels by dividing the count rate by the total static efficiency and correcting for the physical area of the detector plus background. The MDC for static surface activity measurements was calculated using the following equation:

3 + (4.65)

=

Where:

B = background in time interval, T (1 min)

T = count time (min) used for field instruments

= total efficiency = x (instrument efficiency x source efficiency)

G = geometry correction factor (1.26)

The static MDC for beta measurements is 640 dpm/100 cm2 assuming all activity is from Co-60 and an instrument background of 275 cpm. The static MDC for beta measurements is 270 dpm/100 cm2 assuming all activity is from Cs-137 and an instrument background of 275 cpm. The static MDC for SONGS Unit 2 and 3 ISA Confirmatory Survey Report D-3 5365-SR-01-0

alpha measurements is 80 dpm/100 cm2 assuming all activity is from Pu-239/240 and an instrument background of 2 cpm.

D.4. RADIOLOGICAL ANALYSIS D.4.1 GAMMA SPECTROSCOPY Following sample preparation, a portion was sealed in a size appropriate container. The quantity placed in the container was chosen to reproduce the calibrated counting geometry. Net material weights were determined, and the samples were counted using intrinsic, high-purity, geranium detectors coupled to a pulse-height analyzer system. Background and Compton stripping, peak search, peak identification, and concentration calculations were performed using computer capabilities inherent in the analyzer system. Results for the requested radionuclides of concern were provided in units of picocuries per gram (pCi/g).

D.4.2 RADIOACTIVE STRONTIUM ANALYSIS Strontium-90 (Sr-90) concentrations are quantified by total sample dissolution followed by radiochemical separation and are counted on a low background liquid scintillation counter. Soil samples are dissolved by a combination of potassium hydrogen fluoride and pyrosulfate fusions.

Water samples are concentrated, and strontium is separated as a sulfate with further purification.

The sulfate-salts are dissolved in ethylenediaminetetraacetic acid (EDTA). The strontium is separated from residual calcium by re-precipitating strontium sulfate from the EDTA. Strontium sulfate is separated from barium by complexing the strontium sulfate in EDTA and precipitating barium sulfate at a pH of 6. The complexed strontium is precipitated as the sulfate and redissolved in EDTA. The yttrium-90 (Y-90) is allowed to in grow, and the final sample is counted via Cherenkov counting on a liquid scintillation counter. The typical MDC for a 200-minute count time using this procedure is approximately 1 pCi/g for a 1-gram sample.

D.4.3 ALPHA SPECTROMETRY Volumetric samples are dissolved by a combination of potassium hydrogen fluoride and pyrosulfate fusions. Particle samples previously subjected to a pyrosulfate dissolution as described in D.4.2, and a measured quantity of the total dilution are treated with an additional pyrosulfate fusion in a new glass beaker. The fusion cakes are dissolved, and all alpha emitters are co-precipitated on barium SONGS Unit 2 and 3 ISA Confirmatory Survey Report D-4 5365-SR-01-0

sulfate. The barium sulfate is re-dissolved, and the contaminants of concern are separated from the other actinides by either oxidation/precipitations or extraction chromatography utilizing Eichrom Technologies resins, co-precipitated with neodymium fluoride, and analyzed using passivated implanted planar silicon detectors, alpha spectrometers, and multichannel analyzers. The typical MDC for samples with a quantity of 1 (total or grams) is approximately 0.02 pCi/sample or pCi/g.

D.4.4 H-3 ANALYSIS Tritium (H-3) analysis for soil samples are performed using custom distillation glassware and counted by liquid scintillation. The H-3 distillate is collected and mixed with ultima gold LLT (UGLLT) liquid scintillation cocktail. The typical MDC for H-3 for a 100-minute count time using this procedure is approximately 2 pCi/g.

D.4.5 NI-63 ANALYSIS Soil samples are spiked with a nickel (Ni) and cobalt carrier and digested with a mixture of nitric and hydrochloric acids or fused with potassium fluoride/pyrosulfate fusions. Unwanted elements, such as iron and cobalt, are removed via anion exchange chromatography. Nickel is then further separated from the potential interfering elements using a dimethyl glyoxime (DMG) and buffered ammonium citrate. The purified nickel is dissolved in a dilute nitric acid solution, UGLLT liquid scintillation cocktail is added, and the Ni-63 activity is determined via liquid scintillation counting.

The typical MDC for a 1-gram sample and 100-minute count time using this procedure is approximately 2 pCi/g.

D.4.6 TC-99 ANALYSIS Technetium-99 in soil is analyzed by using a sodium hydroxide fusion. The fusion cake is dissolved and passed through a TEVA disk to concentrate and purify the sample. The rinsed TEVA disk is added to a liquid scintillation vile and dissolved with UGLLT and counted on a liquid scintillation counter. The typical MDC for this procedure is approximately 0.4 pCi/g.

D.4.7 DETECTION LIMITS Each RESL analytical result is accompanied by its total propagated uncertainty expressed at one standard deviation. All results that do not pass through zero when their standard deviation is multiplied by two and then added and subtracted to the result are considered statistically positive at SONGS Unit 2 and 3 ISA Confirmatory Survey Report D-5 5365-SR-01-0

the 95% confidence interval. Because of variations in background levels, measurement efficiencies, and contributions from other radionuclides in samples, the detection limits differed from sample to sample and instrument to instrument.

SONGS Unit 2 and 3 ISA Confirmatory Survey Report D-6 5365-SR-01-0

Retrieved from "https://kanterella.com/w/index.php?title=ML23240A537&oldid=3617369"