ML20027A238

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Independent Confirmatory Survey Summary and Results for the Spent Fuel Pool and Transfer Canal at the Zion Nuclear Power Station, Zion, Illinois
ML20027A238
Person / Time
Site: Zion  File:ZionSolutions icon.png
Issue date: 01/24/2020
From: Altic N
Oak Ridge Institute for Science & Education
To: John Hickman
Division of Decommissioning, Uranium Recovery and Waste Programs
Hickman J
References
Download: ML20027A238 (41)
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100 ORAU Way

  • Oak Ridge
  • TN 37830
  • orise.orau.gov January 24, 2020 Mr. John Hickman U.S. Nuclear Regulatory Commission Office of Nuclear Material Safety and Safeguards Division of Decommissioning, Uranium Recovery, and Waste Programs Reactor Decommissioning Branch, Mail Stop: T8F5 11545 Rockville Pike Rockville, MD 20852

SUBJECT:

INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE SPENT FUEL POOL AND TRANSFER CANAL AT THE ZION NUCLEAR POWER STATION, ZION, ILLINOIS DOCKET NOs. 50-295 and 50-304; RFTA 18-004 DCN 5271-SR-05-0

Dear Mr. Hickman:

The Oak Ridge Institute for Science and Education (ORISE) is pleased to provide the attached report detailing the independent confirmatory survey activities associated with the spent fuel pool and transfer canal at the Zion Nuclear Power Station in Zion, Illinois. This report provides the summary and results of ORISE on-site activities performed during the period of July 9-12, 2018.

Please feel free to contact me at 865.574.6273 or Erika Bailey at 865-576-6659 if you have any questions.

Sincerely, Nick A. Altic, CHP Health Physicist/Project Manager ORISE NAA:tb Attachment Electronic distribution:

K. Conway, NRC D. Hagemeyer, ORISE S. Pittman, ORISE R. Edwards, NRC E. Bailey, ORISE File/5271 B. Lin, NRC

INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE SPENT FUEL POOL AND TRANSFER CANAL AT THE ZION NUCLEAR POWER STATION ZION, ILLINOIS N. A. Altic, CHP and S. T. Pittman, Ph.D.

ORISE FINAL REPORT Prepared for the U.S. Nuclear Regulatory Commission January 2020 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.

Zion Spent Fuel Pool and Transfer Canal 5271-SR-05-0 INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE SPENT FUEL POOL AND TRANSFER CANAL AT THE ZION NUCLEAR POWER STATION, ZION, ILLINOIS FINAL REPORT Prepared by N. A. Altic, CHP And S. T. Pittman, Ph.D.

January 2020 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 number 31310018N0014 between the NRC and the U.S. Department of Energy (DOE).

ORISE is managed by Oak Ridge Associated Universities under DOE contract number DE-SC0014664.

Zion Spent Fuel Pool and Transfer Canal 5271-SR-05-0 INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE SPENT FUEL POOL AND TRANSFER CANAL AT THE ZION NUCLEAR POWER STATION, ZION, ILLINOIS Prepared by:

Date:

1/24/2020 N. A. Altic, CHP, Health Physicist/Project Manager ORISE Prepared by:

Date:

1/24/2020 S. T. Pittman, Ph.D., Health Physicist ORISE Reviewed by:

Date:

1/24/2020 P. H. Benton, Quality Manager ORISE Reviewed by:

Date:

1/24/2020 W. F. Smith, Senior Chemist ORISE Reviewed and approved for release by:

Date:

1/24/2020 E. N. Bailey, Survey and Technical Projects Group Manager ORISE FINAL REPORT JANUARY 2020

Zion Spent Fuel Pool and Transfer Canal i

5271-SR-05-0 CONTENTS FIGURES........................................................................................................................................................... ii TABLES.............................................................................................................................................................. ii ACRONYMS.................................................................................................................................................... iii EXECUTIVE

SUMMARY

.............................................................................................................................. v

1. INTRODUCTION....................................................................................................................................... 1
2. SITE DESCRIPTION................................................................................................................................. 2
3. DATA QUALITY OBJECTIVES............................................................................................................. 4 3.1 STATE THE PROBLEM........................................................................................................................ 4 3.2 IDENTIFY THE DECISION/OBJECTIVE........................................................................................... 5 3.3 IDENTIFY INPUTS TO THE DECISION/OBJECTIVE....................................................................... 5 3.3.1 Radionuclides of Concern and Release Guidelines.............................................................. 6 3.4 DEFINE THE STUDY BOUNDARIES................................................................................................. 8 3.5 DEVELOP A DECISION RULE........................................................................................................... 8 3.6 SPECIFY LIMITS ON DECISION ERRORS......................................................................................... 9 3.7 OPTIMIZE THE DESIGN FOR OBTAINING DATA........................................................................ 10
4. PROCEDURES.......................................................................................................................................... 10 4.1 REFERENCE SYSTEM....................................................................................................................... 10 4.2 SURFACE SCANS............................................................................................................................... 10 4.3 MEASUREMENT/SAMPLING LOCATION DETERMINATION....................................................... 10 4.4 ISOCS MEASUREMENTS................................................................................................................. 11 4.5 VOLUMETRIC SAMPLING................................................................................................................ 12
5. SAMPLE ANALYSIS AND DATA INTERPRETATION............................................................... 12
6. FINDINGS AND RESULTS................................................................................................................... 13 6.1 SURFACE SCANS............................................................................................................................... 13 6.2 SFP IN SITU GAMMA SPECTROMETRY MEASUREMENTS........................................................... 13 6.3 ROC CONCENTRATIONS IN CONCRETE SAMPLES..................................................................... 15
7.

SUMMARY

AND CONCLUSIONS...................................................................................................... 17

8. REFERENCES........................................................................................................................................... 18 APPENDIX A: FIGURES APPENDIX B: DATA TABLES APPENDIX C: SURVEY AND ANALYTICAL PROCEDURES APPENDIX D: MAJOR INSTRUMENTATION

Zion Spent Fuel Pool and Transfer Canal ii 5271-SR-05-0 FIGURES Figure 2.1. ZNPS Overview (adapted from ZS 2018)................................................................................. 2 Figure 2.2. ZNPS Spent Fuel Pool/Transfer Canal (Google Earth)......................................................... 3 Figure 4.1. Sample Size Determination Using VSP.................................................................................... 11 Figure 6.1. Q-Q Plot for the Spent Fuel Pool/Transfer Canal Floor and Lower Walls....................... 13 Figure 6.2. Comparison of FSS Data and ORISE Confirmatory Mean Concentrations and Uncertainties for Gamma-emitting Radionuclides in the Spent Fuel Pool............................................. 15 TABLES Table 3.1. ZNPS Confirmatory Survey Decision Process........................................................................... 5 Table 3.2. ZNPS Basement Surfaces DCGLs............................................................................................... 6 Table 6.1. Summary of Spent Fuel Pool Confirmatory Random In Situ Gamma Spectrometry Measurements................................................................................................................................................... 14 Table 6.2. Summary of Analytical Results from Eight Spent Fuel Pool Random Volumetric Concrete Samples.............................................................................................................................................................. 16 Table 6.3. Spent Fuel Pool Concrete Sample Results by Analytical Method.......................................... 17

Zion Spent Fuel Pool and Transfer Canal iii 5271-SR-05-0 ACRONYMS AA alternate action cm centimeter cm3 cubic centimeter(s) cpm counts per minute DCGL derived concentration guideline level DCGLBC Base Case DCGL DCGLOp Operational DCGL DOE U.S. Department of Energy DQO data quality objective DS decision statement EPA U.S. Environmental Protection Agency Exelon Exelon Generation Company FOV field of view FSS final status survey g/cm3 grams per cubic centimeter HDT hard-to-detect HPGe high-purity germanium ISOCS In Situ Object Counting System LTP license termination plan m

meter m2 square meter(s)

MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDC minimum detectable concentration MeV mega electron volt mrem/yr millirem per year NaI sodium iodide NaI(TI) thallium doped sodium iodide NIST National Institute of Standards and Technology NRC U.S. Nuclear Regulatory Commission ORAU Oak Ridge Associated Universities ORISE Oak Ridge Institute for Science and Education pCi/g picocuries per gram pCi/m2 picocuries per square meter PSQ principal study question Q-Q plot quantile-quantile plot ROC radionuclide of concern SA surface area SD standard deviation SFP spent fuel pool SOF sum of fractions SU survey unit TAP total absorption peak TEDE total effective dose equivalent TSD technical support document UCL upper confidence level UCL95 95% upper confidence level

Zion Spent Fuel Pool and Transfer Canal iv 5271-SR-05-0 ACRONYMS (Continued)

VSP Visual Sample Plan WWTF Waste Water Treatment Facility ZNPS Zion Nuclear Power Station ZS ZionSolutions, LLC

Zion Spent Fuel Pool and Transfer Canal v

5271-SR-05-0 INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE SPENT FUEL POOL AND TRANSFER CANAL AT THE ZION NUCLEAR POWER STATION, ZION, ILLINOIS EXECUTIVE

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) requested that the Oak Ridge Institute for Science and Education (ORISE) perform confirmatory survey activities of the remaining structures and surfaces of the spent fuel pool (SFP) and transfer canal at the Zion Nuclear Power Station (ZNPS).

Confirmatory survey activities were conducted during the period of July 9-12, 2018, and consisted of surface scans, in situ gamma spectrometry measurements, and volumetric concrete sampling.

All individual confirmatory measurements, by both in situ measurements and volumetric samples, were below the applicable Operational derived concentration guideline level (DCGLOp), and, therefore, the applicable Base Case DCGL (DCGLBC). ORISE did not identify issues that would preclude final status survey (FSS) data for demonstrating compliance with the release criteria.

Zion Spent Fuel Pool and Transfer Canal 1

5271-SR-05-0 INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE SPENT FUEL POOL AND TRANSFER CANAL AT THE ZION NUCLEAR POWER STATION, ZION, ILLINOIS

1. INTRODUCTION The Zion Nuclear Power Station (ZNPS) consists of two reactors, Units 1 and 2, which operated commercially from 1973 to 1997 and 1974 to 1996, respectively. Cessation of nuclear operations was certified in 1998 after both reactor units were defueled and the fuel assemblies had been placed in the spent fuel pool (SFP). Both units were then placed in safe storage pending the commencement of site decommissioning and dismantlement. In 2010, the U.S. Nuclear Regulatory Commission (NRC) operating license was transferred from Exelon Generation Company (Exelon) to ZionSolutions, LLC (ZS) to allow the physical decommissioning process that began in 2010 and is expected to be completed within 10 years. The end-state and primary decommissioning objective at ZNPS is the transfer of all spent nuclear fuel to the independent spent fuel storage installation and to reduce residual radioactivity levels below the criteria specified in 10 CFR 20.1402, permitting release of the site for unrestricted use. Upon successful completion of the decommissioning activities, control and responsibility for the site will be transferred back to Exelon and the independent spent fuel storage installation maintained under Exelons Part 50 license (EC 2015).

ZSs decommissioning commitments were that, all above-grade structures, with minor exceptions, would be demolished. Structures below the 588-foot elevation (referenced from mean sea level),

consisting of primarily exterior subgrade walls and floors, would remain. These basement structures would be backfilled as part of the final site restoration. In order to demonstrate compliance with the release criteria in 10 CFR 20.1402, ZS would implement final status survey (FSS) activities of remaining basement structures along with associated embedded piping and penetrations, buried piping, and surface and subsurface soil. FSS methodologies are outlined in Chapter 5 of ZSs license termination plan (LTP) (ZS 2018). NRC issued license amendments 178 and 191 to approve ZSs LTP in September of 2018 (NRC 2018). The primary FSS method utilized for basement structure survey units (SUs) was in situ gamma measurements using a portable, high-resolution gamma spectrometer. FSS methods were based on methods outlined in the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) (NRC 2000).

Zion Spent Fuel Pool and Transfer Canal 2

5271-SR-05-0 NRC requested that the Oak Ridge Institute for Science and Education (ORISE) perform confirmatory survey activities of the remaining structures and surfaces of the SFP/transfer canal at ZNPS. This report summarizes the confirmatory survey activities and results for these areas.

2. SITE DESCRIPTION ZNPS is located in Lake County, Illinois, on the easternmost portion of the city of Zion. It is approximately 64 kilometers (40 miles) north of Chicago, Illinois, and 68 kilometers (42 miles) south of Milwaukee, Wisconsin. The owner-controlled site is composed of approximately 134 hectares (331 acres) and is situated between the northern and southern parts of Illinois Beach State Park on the western shore of Lake Michigan (EC 2015 and ZS 2018). Figure 2.1 provides an overview of ZNPS. The site and its surrounding environs is relatively flat with the elevation of the developed portion of the site at approximately 591 feet above mean sea level. For reference, the elevation of Lake Michigan, which bounds the site on the east, is approximately 577.4 feet at low water level (ZS 2018).

Figure 2.1. ZNPS Overview (adapted from ZS 2018)

Zion Spent Fuel Pool and Transfer Canal 3

5271-SR-05-0 The SFP housed the spent fuel assemblies after removal from the reactor. The SFP was located between the Unit 1 and Unit 2 Containment Buildings and immediately to the west of the Auxiliary Building. The SFP has been removed as part of decommissioning activities, leaving only the concrete that was below the bottom of the SFP. The historical site assessment noted two instances where the SFP overflowed and also noted leakage around the SFP weld liner.

Figure 2.2 displays the licensed area and indicates the approximate location of the SFP relative to the Containment Buildings.

Figure 2.2. ZNPS Spent Fuel Pool/Transfer Canal (Google Earth)

Spent Fuel Pool/Transfer Canal Containment Buildings

Zion Spent Fuel Pool and Transfer Canal 4

5271-SR-05-0

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 provide a formalized method for planning radiation surveys, improving survey efficiency and effectiveness, and ensuring that the type, quality, and quantity of data collected are adequate for the intended decision applications.

The seven steps of the DQO process are 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 Confirmatory survey DQOs were originally presented in ORISE 2018 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. Prior to the confirmatory site visit, ZS was in the process of dismantling remaining structures and remediating remaining land areas. As part of this process, ZS conducted an FSS to demonstrate compliance with NRCs license termination criteria specified in 10 CFR 20.1402. To this end, NRC requested that ORISE perform confirmatory activities of the SFP to provide NRC with independent confirmatory data for NRCs consideration in the evaluation of the FSS. The problem statement was formulated as follows:

Confirmatory surveys are necessary to generate independent radiological data for NRCs consideration in the evaluation of the FSS design, implementation, and results for demonstrating compliance with the release criteria.

Zion Spent Fuel Pool and Transfer Canal 5

5271-SR-05-0 3.2 IDENTIFY THE DECISION/OBJECTIVE The second step in the DQO process identified the principal study questions (PSQs) and alternate actions (AAs), developed a decision statement (DS), and organized multiple decisions, as appropriate. This was done by specifying AAs that could result from a yes response to the PSQ and combining the PSQ and AAs into a DS. Table 3.1 presents the confirmatory survey decision process.

Table 3.1. ZNPS Confirmatory Survey Decision Process Principal Study Question Alternate Actions Do confirmatory survey results agree with the final radiological survey data for the SFP?

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 the NRC-approved decommissioning criteria, and/or that statistical sample population examination/assessment conditions were met.

No:

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.

Decision Statement Confirmatory survey results did/did not identify anomalous results or other conditions that preclude the FSS data from demonstrating compliance with the release criteria.

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 would meet data requirements. For this effort, information inputs included the following:

  • Concrete characterization and FSS data for the SFP
  • Derived concentration guideline levels (DCGLs), discussed in subsection 3.3.1

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5271-SR-05-0

  • ORISE confirmatory survey results, including surface radiation scans, direct surface activity measurements, and in situ gamma spectroscopy measurements with an In Situ Object Counting System (ISOCS)
  • ORISE volumetric analytical results for concrete samples 3.3.1 Radionuclides of Concern and Release Guidelines The primary radionuclides of concern (ROCs) identified for ZNPS are beta-gamma emitters fission and activation productsresulting from reactor operations. At ZNPS, there are four distinct source terms: basement structures, soils, buried piping, and groundwater. Furthermore, basement structures are composed of four structural source terms: surfaces, embedded piping, penetrations, and fill. ZS developed site-specific DCGLs that correspond to a residual radioactive contamination level above background, which could result in a total effective dose equivalent (TEDE) of 25 millirem per year (mrem/yr) to an average member of the critical group. These DCGLsdefined in ZSs LTP as Base Case DCGLs (DCGLBCs)are radionuclide specific and independently correspond to a TEDE of 25 mrem/yr for each source term.

In order to ensure that total dose from all source terms is less than the NRC-approved release criteria, the DCGLBCs were further reduced to Operational DCGLs (DCGLOps). The DCGLOps were scaled to an expected dose from prior investigations and were used for remediation and FSS design purposes. The initial suite of radionuclides present at ZNPS was reduced based on an insignificant dose contribution from a number of radionuclides. The DCGLBCs and DCGLOps, accounting for insignificant dose contributors, for the basement structure source termsexcluding fill material, are presented in Table 3.2.

Table 3.2. ZNPS Basement Surfaces DCGLsa ROC Auxiliary Building Containment SFP/Transfer Canal Turbine Building Crib House

/Forebay WWTF Above 565 feet Under-vessel Floors &

Wallsb Circ Water Discharge Tunnel Base Case DCGLs (pCi/m2)

H-3 5.30E+08 2.38E+08 2.38E+08 1.29E+08 1.93E+08 1.71E+07 Co-60 3.04E+08 1.57E+08 1.57E+08 7.03E+07 5.52E+07 2.83E+07 Ni-63 1.15E+10 4.02E+09 4.02E+09 2.18E+09 3.25E+09 2.89E+08 Cs-134 2.11E+08 3.01E+07 3.01E+07 1.59E+07 2.13E+07 2.31E+06

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5271-SR-05-0 Table 3.2. ZNPS Basement Surfaces DCGLsa ROC Auxiliary Building Containment SFP/Transfer Canal Turbine Building Crib House

/Forebay WWTF Above 565 feet Under-vessel Floors &

Wallsb Circ Water Discharge Tunnel Cs-137 1.11E+08 3.94E+07 3.94E+07 2.11E+07 2.96E+07 2.93E+06 Sr-90 9.98E+06 1.43E+06 1.43E+06 7.74E+05 1.16E+06 1.03E+05 Eu-152 6.47E+08 3.66E+08 3.66E+08 1.62E+08 1.23E+08 7.55E+07 Eu-154 5.83E+08 3.19E+08 3.19E+08 1.43E+08 1.12E+08 5.74E+07 Operational DCGLs (pCi/m2)

H-3 1.71E+08 3.25E+07 2.37E+08 4.98E+07 1.10E+07 5.39E+07 7.43E+07 3.28E+06 Co-60 9.81E+07 2.15E+07 1.56E+08 3.28E+07 5.98E+06 2.94E+07 2.13E+07 5.43E+06 Ni-63 3.71E+09 5.50E+08 4.00E+09 8.41E+08 1.85E+08 9.11E+08 1.25E+09 5.55E+07 Cs-134 6.81E+07 4.12E+06 2.99E+07 6.30E+06 1.35E+06 6.65E+06 8.20E+06 4.44E+05 Cs-137 3.58E+07 5.39E+06 3.92E+07 8.24E+06 1.79E+06 8.82E+06 1.14E+07 5.63E+05 Sr-90 3.22E+06 1.96E+05 1.42E+06 2.99E+05 6.58E+04 3.24E+05 4.47E+05 1.98E+04 Eu-152 2.09E+08 5.00E+07 3.64E+08 7.66E+07 1.38E+07 6.77E+07 4.74E+07 1.45E+07 Eu-154 1.88E+08 4.36E+07 3.17E+08 6.67E+07 1.22E+07 5.98E+07 4.31E+07 1.10E+07 aRecreated from ZS 2018 bThe DCGLOps for floors and walls will be applied to surfaces in the Circulating Water Intake Pipe and Circulating Water Discharge Pipe SFP = spent fuel pool WWTF = Waste Water Treatment Facility Because each of the individual DCGLBCs represent a separate radiological dose, the sum-of-fractions (SOF) approach must be used to evaluate the total dose from the SU and demonstrate compliance with the dose limit. SOF calculations were performed as follows:

=

Cmean,j DCGLBC,j

=1

+

,x Eq. 3-1 Where:

Cmean,j is the mean concentration of ROC j CElv,j is an elevated area of ROC j DCGLBC,j is the Base Case DCGL for ROC j SASU is the adjusted surface area (SA) of the FSS unit SAElv. is the SA of the elevated measurement.

Zion Spent Fuel Pool and Transfer Canal 8

5271-SR-05-0 Per Section 5.5.4 of the LTP, areas of elevated activityfor building surfaceswere defined as any area identified by measurement/sample (systematic or judgmental) that exceeds the DCGLOp, but is less than the DCGLBC. Any area that exceeds the DCGLBC required remediation. ORISE measurements were compared to the DCGLBC. Note that gross concentrations are considered here for conservatism.

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. Confirmatory surveys were conducted for the SFP during the period of July 9-12, 2018.

Temporal limitations prevented ORISE from achieving 100% areal coverage with the in situ measurements, as achieved with the FSS measurements. The SFP survey unit identification number is 0302AF, as assigned by ZS.

3.5 DEVELOP A DECISION RULE The fifth step in the DQO process specified appropriate population parameters (e.g., mean, median), confirmed action levels are above detection limits, and developed an ifthen decision rule statement. Decision rules for this survey were based on independent scan surveys, in situ gamma spectrometry measurements, and concrete sample results to assess whether there is a statistical bias relative to the FSS data. Typically, decision rules are based on a statistical comparison of the ORISE survey data and the FSS data using an appropriate test. However, the difference in the ORISE and FSS sample sizes is significant. For example, 76 FSS in situ gamma spectroscopy measurements were collected in the SFP SU, whereas ORISE collected only 14 random measurements. The approximately 5:1 sample size ratio significantly reduces the statistical power to detect a difference between the two sample groups. Therefore, alternative assessment methods were employed.

The parameter of interest for the survey area was the mean ROC concentration and associated confidence level. The mean and associated uncertainty for the confirmatory measurements was compared directly to that of the FSS data, and agreement between the two sample sets was evaluated by the overlapping of confidence intervals. The SOF of the confirmatory survey data was also compared directly to the DCGLBC. The aforementioned information was combined to formulate the decision rule for the SFP, which was stated as follows:

Zion Spent Fuel Pool and Transfer Canal 9

5271-SR-05-0 If the mean ROC concentration confidence intervals of the confirmatory and FSS sample populations overlap and confirmatory measurements are below criteria, then conclude that the confirmatory survey data agrees with the FSS data; otherwise, perform further evaluation(s) and provide technical comments to NRC.

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 both during planning and during data quality assessment and were based on two orders of control.

The first order of control was related to sample size, which impacts the degree to which the estimated sample mean is bound. Visual Sample Plan (VSP), version 7.9, was used to determine the confirmatory survey sample size using the FSS/characterization data as planning inputs. The constraint on the estimated mean was not larger than the difference between the DCGL and the reported FSS data mean. Numerically, the constraint on the estimated mean was represented by (half-width of the estimated mean 1.00 - FSS Mean SOF). The confirmatory survey mean was estimated at the 95% confidence level. (See section 4.3 of this report for additional details.)

As stated in Section 3.4, temporal limitations prevented ORISE from achieving 100% areal coverage of the SFP with the in situ measurements. Thus, the presence or absence of all potential locations not represented by the arithmetic mean of the confirmatory data set could not be identified and accounted for using Equation 3-1. Therefore, the 95% upper confidence level (UCL95) of the mean was used for the SOF calculations. The UCL accounts for the uncertainty in the estimate of the mean due to sampling error; therefore, the probability of a Type I decision error is minimized (i.e., incorrectly concluding that the ROC concentration is less than DCGLBC).

The second order of control was to optimize the analytical minimum detectable concentrations (MDCs) with respect to ORISE sample count times for field and laboratory measurements.

Measurement MDCs for the in situ gamma spectrometry measurements were less than 50% of the applicable guidelines presented in Section 3.3.1. Nominal MDCs for laboratory instrumentation were sufficient for the evaluation of ROC concentrations in volumetric media and subsequent decision making.

Zion Spent Fuel Pool and Transfer Canal 10 5271-SR-05-0 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 upon alternatives, and document requisite details. Specific survey procedures are presented in Section 4.

4. PROCEDURES The ORISE survey team conducted independent confirmatory survey activities, including surface scans, in situ gamma spectrometry measurements, and volumetric sampling activities, within the accessible survey area specifically requested by NRC. Survey activities were conducted in accordance with the Oak Ridge Associated Universities (ORAU) Radiological and Environmental Survey Procedures Manual and the ORAU Environmental Services and Radiation Training Quality Program Manual (ORAU 2016a and ORAU 2016b).

4.1 REFERENCE SYSTEM ORISE used specific X-Y coordinates from the southwest corner of the respective SU floors and lower left corner of walls to reference measurement and sampling locations that were documented on detailed survey maps.

4.2 SURFACE SCANS Thallium doped sodium iodide (NaI[Tl]) detectors were used to evaluate direct gamma radiation levels on basement surfaces. Accessible SFP surfaces received high-density scan coverage.

All detectors were coupled to Ludlum Model 2221 ratemeter-scalers with audible indicators and also were coupled to data-loggers to electronically record all scanning data. Locations of elevated response that are audibly distinguishable from localized background levels, suggesting the presence of residual contamination, were marked for further investigation with in situ gamma spectrometry measurements and/or volumetric sampling.

4.3 MEASUREMENT/SAMPLING LOCATION DETERMINATION Measurement locations were determined both randomly and judgmentally. VSP was used to assess the sample size required for decision making and to randomly place locations throughout the survey

Zion Spent Fuel Pool and Transfer Canal 11 5271-SR-05-0 area. The FSS in situ gamma spectroscopy data indicated a mean SOF of 0.06 and a corresponding standard deviation of 0.06. Using these values as inputs, VSP calculated that seven random samples were needed to conclude that the estimated mean SOF falls between 0 and 0.12 at the 95%

confidence level using simple random sampling. Figure 4.1 illustrates the sample size determination using VSP.

Figure 4.1. Sample Size Determination Using VSP To improve confidence, 16 random in situ measurements were planned; however, only 14 were collected because of access limitations. Eight concrete core samples were randomly collected from the 16 in situ measurement locations. Two judgmental in situ measurements were also collected from regions of interest identified during survey activities.

4.4 ISOCS MEASUREMENTS In situ gamma spectrometry measurements were performed at the randomly selected locations planned and mapped using VSP. Judgmental measurement locations were selected based on professional judgement. The detector was positioned and collimated such that the field of view (FOV) was approximately 28 square meters (m2) to coincide with the FSS measurement FOV.

Measurements were performed using a portable Broad-Energy High-Purity Germanium (HPGe) detector. Data acquisition was performed via Canberras Genie 2000 software. Efficiency curves

Zion Spent Fuel Pool and Transfer Canal 12 5271-SR-05-0 encompassing the applicable radionuclidesfor the measurement geometry were generated using Canberras ISOCS calibration software. The efficiency curves were modeled using ZSs ROC depth profile that was based on applicable characterization data. Specific modeling parameters are discussed in Appendix C.

4.5 VOLUMETRIC SAMPLING Concrete samples were collected from randomly-and judgmentally-selected locations using a concrete hole saw and an electric drill. Samples were collected from a depth of up to 15 centimeters (cm) (6 inches), or until refusal, and sample depth noted at each location.

Concrete cores were sectioned into 5-cm (approximately 2-inch) increments, for a maximum of three increments, before submitting for laboratory analysis. The increments were analyzed individually starting from the top portion (i.e., the 0- to 5-cm portion). In the event that radionuclides in the top increment were not detected above the analytical MDC, subsequent increments were not analyzed.

5. SAMPLE ANALYSIS AND DATA INTERPRETATION Samples and data collected on site were transferred to the ORISE facility for analysis and interpretation. Sample custody was transferred to the Radiological and Environmental Analytical Laboratory in Oak Ridge, Tennessee. Sample analyses were performed in accordance with the ORAU Radiological and Environmental Analytical Laboratory Procedures Manual (ORAU 2017). Concrete samples were analyzed by gamma spectrometry for gamma-emitting fission and activation products.

Portions of each concrete sample were processed by wet chemistry and material oxidation and then analyzed for Sr-90, Ni-63, and H-3 by low-background proportional or liquid scintillation analyzer counting, as applicable, after separation. Volumetric sample results in units of picocuries per gram (pCi/g) were converted to units of pCi/m2 based on the concrete sample depth. Measurement results from the in situ gamma spectrometry measurements were reported in units of pCi/m2.

ProUCL, version 5.1, was used to calculate the UCL95 for both the confirmatory and FSS data sets.

The mean ROC concentration and associated 95% confidence level were plotted for direct comparison.

Zion Spent Fuel Pool and Transfer Canal 13 5271-SR-05-0

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

6.1 SURFACE SCANS Overall, NaI detector scan responses ranged from approximately 4,100 to 94,000 counts per minute (cpm) for the floor of the SFP and transfer canal and 6,500 to 15,000 cpm for the lower walls.

Figure 6.1 presents the quantile-quantile (Q-Q) plot for these scan data. The sharp increase in NaI detector responses, as indicated by the steep curve for the floor in Figure 6.1, was due to gamma shine from the Auxiliary Building. The surveyors noted an increase in detector response along the Auxiliary Building/SFP SU boundary.

Figure 6.1. Q-Q Plot for the Spent Fuel Pool/Transfer Canal Floor and Lower Walls 6.2 SFP IN SITU GAMMA SPECTROMETRY MEASUREMENTS Sixteen in situ gamma spectrometry measurementsincluding 14 random and two judgmental measurementswere collected from the SFP. Individual in situ gamma spectrometry measurements collected from this area are presented in Table B.1 in Appendix B. Measurement locations from this area are depicted in Figure A.1 in Appendix A. Table 6.1 provides a summary of the confirmatory

Zion Spent Fuel Pool and Transfer Canal 14 5271-SR-05-0 random/systematic in situ gamma spectrometry measurements. Randomly-planned locations 5271SFP-11 and 5271SFP-16 were not collected because of access limitations. (The personnel bridge into the containment buildings prevented positioning the HPGe detector to measure these locations.) Two judgmental in situ measurement locations (5271SFP-17J and 5271SFP-18J) were selected along the western boundary of the SFP. These locations were not selected based on surface scan results, but, rather, because a random/systematic sample was not collected from this area.

Table 6.1. Summary of Spent Fuel Pool Confirmatory Random In Situ Gamma Spectrometry Measurements ROC Parameter (pCi/m2)

Fractionb Mean Median SD Min Max UCL95a Op.

BC Co-60 4.07E+04 3.20E+04 2.89E+04 5.03E+03 1.16E+05 7.43E+04

<0.01 <0.01 Cs-134 6.16E+03 2.78E+03 1.84E+04

-1.41E+04 6.36E+04 2.75E+04

<0.01 <0.01 Cs-137 3.46E+04 2.37E+04 3.86E+04

-4.76E+03 1.50E+05 7.95E+04 0.01

<0.01 SOFc 0.02

<0.01 aUCL is based on the Chebyshev Inequality bOp. represents the UCL95 divided by the DCGLOp; BC represents the UCL95 divided by the DCGLBC cDiscrepancy in summation is due to rounding SD = standard deviation The only gamma-emitting radionuclides identified above their respective MDCs were Cs-137 and Co-60. The maximum measurement for each ROC was less than the respective DCGLOp and, therefore, less than the respective DCGLBC. The average SOF in this SU for gamma-emitting radionuclides is less than 0.01 based on the DCGLBC (or 0.02 when based on the DCGLOp), which is less than the SOF for gamma-emitting radionuclides (0.03) presented in the FSS data.

Gamma-emitting ROC mean concentrations and their associated uncertainties for both the ORISE and FSS data are plotted in Figure 6.2. The error bars in Figure 6.2 represent the uncertainty in the mean concentration, where the upper end is the UCL95. As indicated in Figure 6.2, all ROC mean concentrations overlap at the 95% confidence level. All mean ROC concentrations determined for the FSS data were high relative to those determined by ORISE. Because the FSS achieved 100%

area coverage, several measurements in the FSS sample population were collected near the Auxiliary Building, which contributed additional counts to the acquired FSS in situ spectrum from radionuclidesspecifically Cs-137contained within the concrete in the Auxiliary Building.

Zion Spent Fuel Pool and Transfer Canal 15 5271-SR-05-0 Figure 6.2. Comparison of FSS Data and ORISE Confirmatory Mean Concentrations and Uncertainties for Gamma-emitting Radionuclides in the Spent Fuel Pool 6.3 ROC CONCENTRATIONS IN CONCRETE SAMPLES Eight volumetric concrete samples were collected from randomly selected in situ gamma spectrometry measurement locations. Table 6.2 presents a summary of the concrete samples collected. Individual results for volumetric concrete samples are provided in Table B.2 in Appendix B. The concrete samples were collected from the in situ gamma spectroscopy measurement locations as indicated in Figure A-1. Concrete sample analytical results were converted to units of pCi/m2 based on a sample depth of 5.08 cmsame as input into the ISOCS modeland a concrete depth of 2.35 grams per cubic centimeter (g/cm3). None of the individual ROC concentrations were above their respective DCGLOp and, therefore, were less than the respective DCGLBC.

Zion Spent Fuel Pool and Transfer Canal 16 5271-SR-05-0 Table 6.2. Summary of Analytical Results from Eight Spent Fuel Pool Random Volumetric Concrete Samples ROC Parameter (pCi/m2)

Mean Median SD Min Max UCL95a Co-60 3.04E+03 1.43E+03 6.47E+03

-2.51E+03 1.79E+04 1.30E+04 Cs-134 1.64E+03 2.15E+03 2.48E+03

-2.15E+03 4.54E+03 5.47E+03 Cs-137 3.12E+03 1.73E+03 3.91E+03 0.00E+00 1.23E+04 9.13E+03 H-3 2.61E+05 1.91E+05 3.46E+05

-1.07E+05 1.05E+06 7.95E+05 Ni-63

-1.15E+04

-1.07E+04 1.92E+04

-3.94E+04 1.31E+04 1.81E+04 Sr-90 5.19E+04 5.43E+04 2.22E+04 2.63E+04 9.31E+04 8.61E+04 aUCL95 is based on the Chebyshev Inequality Table 6.3 provides the UCL95 of the mean ROC concentrations determined by both the volumetric samples and by the in situ gamma spectrometry measurements for the SFP along with the corresponding fractional contribution to the SOF from each ROC. Cs-137 was identified above the analytical MDC in sample M0039. No other concrete samples had the gamma-emitting radionuclides of concern present at concentrations above their respective analytical MDCs. Ni-63 and H-3 were identified above the MDC in samples M0039 and M0040, respectively. As a result of these positive detects, the additional depth increments were analyzed at these locations and neither Ni-63 nor H-3 were identified in the increments representing a 5-to 10-cm (approximately 2-to 6-inch) concrete depth. All sample results for Sr-90 were below the MDC. The mean hard-to-detect (HTD)

(consisting of H-3, Ni-63, and Sr-90) SOF, when compared to respective DCGLBC values, was determined to be 0.02 compared to 0.00 (i.e., less than 0.01) reported by the licensee. The primary difference between the HTD SOF values is due to the difference in MDCs in the respective analysis methods (volumetric sampling for the ORISE confirmatory data versus the surrogate approach for in situ FSS measurements).

Table 6.3 indicates the in situ measurements for the gamma ROC concentrations are biased high relative to the volumetric concentrations. ROC-specific fractions based on the DCGLBC, determined by each analysis method, are agreeable to two decimal places for Co-60, Cs-134, and Cs-137.

Zion Spent Fuel Pool and Transfer Canal 17 5271-SR-05-0 Table 6.3. Spent Fuel Pool Concrete Sample Results by Analytical Method ROC Volumetric Concrete Sample In Situ by ISOCS Concentrationa (pCi/m2)

Fractionb Concentrationa (pCi/m2)

Fractionb Op.

BC Op.

BC Co-60 1.30E+04

<0.01

<0.01 7.43E+04

<0.01 0.01 Cs-134 5.47E+03

<0.01

<0.01 2.75E+04

<0.01

<0.01 Cs-137 9.13E+03

<0.01

<0.01 7.95E+04 0.01

<0.01 Gamma-emitting Fractionc

<0.01

<0.01 0.02

<0.01 H-3 7.95E+05 0.02

<0.01

--d Sr-90 1.81E+04 0.06

<0.01 Ni-63 8.61E+04 0.00

<0.01 SOFc 0.08 0.02 aReported concentration is the UCL95 based on Chebyshev's Inequality bOp. represents the UCL95 divided by the DCGLOp; BC represents the UCL95 divided by the DCGLBC cDiscrepancy in summation is due to rounding dInferred concentration is not applicable

7.

SUMMARY

AND CONCLUSIONS At NRCs request, ORISE conducted confirmatory survey activities at ZNPS during the period of July 9-12, 2018. The survey activities included gamma surface scans, in situ gamma spectrometry measurements at 14 random/systematic and 2 judgmental locations, and volumetric sampling at 8 of the 14 random locations.

All individual confirmatory measurements, by both in situ measurements and volumetric samples, were well below the DCGLOp and, therefore, also were below the DCGLBC. Based on the overlap of confidence intervals and relative mean SOF magnitudes between the confirmatory and FSS data for the in situ gamma spectrometry measurements, ORISE did not identify issues that would preclude the gamma-emitting ROC FSS data for demonstrating compliance with the release criteria. The volumetric concrete cores did not identify HTD (H-3, Sr-90, Ni-63) concentrations above their respective analytical MDC at depths greater than 5 cm (2 inches). Therefore, the volumetric concrete results confirmed that the ISOCS modeling parameters were adequate.

In conclusion, the mean ROC concentrations of the confirmatory and FSS sample populations overlap at the 95% confidence level and confirmatory analytical results are below the applicable criteria. Therefore, based on the parameters established by project-specific DQOs, confirmatory survey data agree with the FSS data.

Zion Spent Fuel Pool and Transfer Canal 18 5271-SR-05-0

8. REFERENCES Canberra 2009. Model S573 ISOCS Calibration Software Technical Reference Manual. Canberra Industries, Inc. Meriden, Connecticut. EC 2015. The Future of Zion. Webpage:

http://www.exeloncorp.com/locations/power-plants/zion-station. Exelon Corporation. Chicago, Illinois. Accessed June 30, 2015.

EC 2015. The Future of Zion. Webpage: http://www.exeloncorp.com/locations/power-plants/zion-station. Exelon Corporation. Chicago, Illinois. Accessed June 30, 2015.

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 Survey and Site Investigation Manual (MARSSIM). NUREG-1575; Revision 1. U.S. Nuclear Regulatory Commission. Washington, D.C. August.

NRC 2018. Letter from J.B. Hickman, USNRC, to J. Sauger, Energy Solutions, RE: Zion Nuclear Power Station, Units 1 and 2 - Issuance of Amendments 191 and 178 for the Licenses to Approve the License Termination Plan. September 28.

ORAU 2014. ORAU Radiation Protection Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. October.

ORISE 2015. Request for Additional information from the Independent Technical Review of Technical Support Document Use of In-Situ Gamma Spectroscopy for Source Term Survey of End State Structures, Zion Nuclear Power Station, Zion, Illinois. 5271-DR-01-0. Oak Ridge Institute for Science and Education.

Oak Ridge, Tennessee. October 14.

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

ORAU 2016b. ORAU Environmental Services and Radiation Training Quality Program Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. November 9.

ORAU 2016c. ORAU Health and Safety Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. January.

ORAU 2017. ORAU Radiological and Environmental Analytical Laboratory Procedures Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. August 24.

ORISE 2018. Project-Specific plan for Confirmatory Survey Activities of the Spent Fuel Pool and Transfer Canal at the Zion Nuclear Power Station. Oak Ridge Institute for Science and Education. Oak Ridge, Tennessee. July 6.

ZS 2018. Zion Station Restoration Project License Termination Plan, Rev. 2. ZionSolutions, LLC. Chicago, Illinois. February 7.

Zion Spent Fuel Pool and Transfer Canal 5271-SR-05-0 APPENDIX A FIGURES

Zion Spent Fuel Pool and Transfer Canal A-1 5271-SR-05-0 Random/Systematic In Situ Random/Systematic In Situ and Concrete Judgmental In Situ ORISE Zion NPS Zion, Illinois SFP Sample Locations Created by: A. Kirthlink October 10, 2018 Y:\\IEAV\\Projects\\5271 Zion NPS Figure A.1. Spent Fuel Pool In Situ Gamma Spectrometry Measurement Locations

Zion Spent Fuel Pool and Transfer Canal 5271-SR-05-0 APPENDIX B DATA TABLES

Zion Spent Fuel Pool and Transfer Canal B-1 5271-SR-05-0 Table B.1. In Situ Gamma Spectrometry Measurements from the Spent Fuel Pool (pCi/m2)

Gamma Spec Measurement IDa Co-60 Cs-134 Cs-137 Result MDC Result MDC Result MDC 5271SFP-01 1.90E+04 2.96E+04 3.86E+03 2.64E+04 1.78E+04 3.24E+04 5271SFP-02 1.86E+04 3.15E+04

-6.30E+03 2.29E+04 2.14E+04 3.00E+04 5271SFP-03 2.58E+04 3.22E+04 1.03E+04 2.09E+04 1.51E+04 2.78E+04 5271SFP-04 1.21E+04 3.02E+04

-2.15E+03 2.47E+04 -4.76E+03 3.00E+04 5271SFP-05 4.63E+04 4.47E+04 1.07E+03 2.97E+04 3.09E+04 1.93E+04 5271SFP-06 3.60E+04 3.51E+04 5.38E+03 2.47E+04 -4.55E+03 3.00E+04 5271SFP-07 2.80E+04 3.78E+04 1.70E+03 2.84E+04 1.86E+04 3.14E+04 5271SFP-08 2.51E+04 3.46E+04

-2.34E+03 2.79E+04 1.85E+04 3.14E+04 5271SFP-09 5.47E+04 4.41E+04

-4.19E+03 2.94E+04 3.98E+04 3.56E+04 5271SFP-10 5.03E+03 2.65E+04 4.84E+03 3.10E+04 2.59E+04 3.29E+04 5271SFP-12 6.29E+04 1.67E+04 2.00E+04 3.46E+04 1.50E+05 2.45E+04 5271SFP-13 1.16E+05 3.59E+04 6.36E+04 1.68E+05 5.90E+04 2.70E+04 5271SFP-14 5.92E+04 1.74E+04 4.58E+03 2.84E+04 6.38E+04 2.08E+04 5271SFP-15 6.07E+04 4.62E+04

-1.41E+04 2.41E+04 3.25E+04 4.06E+04 5271SFP-17Jb 1.88E+04 3.13E+04

-1.45E+04 2.82E+04 8.39E+03 2.74E+04 5271SFP-18Jb 1.99E+04 3.60E+04

-7.22E+03 2.72E+04 -2.08E+03 3.17E+04 aLocations associated with 5271SFP-11 and -16 were inaccessible; therefore, the associated in situ measurements could not be collected.

aJudgmental measurement

Zion Spent Fuel Pool and Transfer Canal B-2 5271-SR-05-0 Table B.2. ROC Concentrations from Eight Spent Fuel Pool Concrete Sample Locations Sample IDa Gamma Spec ID Concentration (pCi/m2)b Co-60 Cs-134 Cs-137 H-3 Sr-90 Ni-63 5271M0039A 5271SFP-15 5.61E+03

-4.78E+02 1.23E+04c 1.31E+05

-3.22E+04 9.31E+04 5271M0039B 5271SFP-15

-1.19E+02

-9.55E+02 2.39E+02 2.75E+05

--d 6.69E+04 5271M0040A 5271SFP-14

-2.51E+03

-2.15E+03 1.19E+03 1.05E+06 3.58E+03 2.63E+04 5271M0040B 5271SFP-14

-1.43E+03 4.54E+03 9.55E+02 2.39E+05 5.97E+04 5271M0041A 5271SFP-08 1.79E+03 4.54E+03 0.00E+00 7.16E+04

-3.94E+04 6.57E+04 5271M0042A 5271SFP-07 1.07E+03 2.87E+03 1.55E+03 2.51E+05

-9.55E+03 5.61E+04 5271M0043A 5271SFP-02 1.79E+03 1.43E+03 4.30E+03 3.22E+05

-2.39E+04 5.73E+04 5271M0044A 5271SFP-05 1.79E+04 2.87E+03 1.91E+03

-1.07E+05 8.36E+03 2.63E+04 5271M0045A 5271SFP-12

-7.16E+02

-4.78E+02 1.31E+03 2.63E+05

-1.19E+04 5.25E+04 5271M0046A 5271SFP-10

-5.97E+02 4.54E+03 2.39E+03 1.07E+05 1.31E+04 3.82E+04 a"A" represents 0 to 5.08 cm increment; "B" represents 5 to 10 cm increment bValues converted to units of pCi/m2 based on a concrete density of 2.35 g/cm3 and sample depth of 5.08 cm.

cBolded values indicate results greater than the analytical MDC in pCi/g before conversion to pCi/m2.

dIndicates analysis not performed

Zion Spent Fuel Pool and Transfer Canal 5271-SR-05-0 APPENDIX C: SURVEY AND ANALYTICAL PROCEDURES

Zion Spent Fuel Pool and Transfer Canal C-1 5271-SR-05-0 C.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 2014, ORAU 2016a, and ORAU 2016c). 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 walkdown 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 the ORAU Radiological and Environmental Survey Procedures Manual (ORAU 2016a) 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.

C.2.

CALIBRATION AND QUALITY ASSURANCE Calibration of all field instrumentation was based on standards/sources, traceable to the 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 2016a)
  • ORAU Environmental Services and Radiation Training Quality Program Manual (ORAU 2016b)
  • ORAU Radiological and Environmental Analytical Laboratory Procedures Manual (ORAU 2017)

The procedures contained in these manuals were developed to meet the requirements of U.S. Department of Energy (DOE) Order 414.1D and the U.S. Nuclear Regulatory Commission (NRC) 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

Zion Spent Fuel Pool and Transfer Canal C-2 5271-SR-05-0 operation is within acceptable statistical fluctuations.

  • Participation in Mixed-Analyte Performance Evaluation Program and Intercomparison Testing Program laboratory quality assurance programs.
  • Training and certification of all individuals performing procedures.
  • Periodic internal and external audits.

C.3 SURVEY PROCEDURES C.3.1 SURFACE SCANS Scans for elevated gamma radiation were performed by passing the detector slowly over the surface.

The distance between the detector and surface was maintained at a minimum. Specific scan minimum detectable concentration (MDCs) for the sodium iodide scintillation detectors (NaI) were not determined as the instruments were used solely as a qualitative means to identify elevated gamma radiation levels in excess of background. Identifications of elevated radiation levels that could exceed the site criteria are determined based on an increase in the audible signal from the indicating instrument.

C.3.2 IN SITU GAMMA SPECTROMETRY MEASUREMENTS Canberras In Situ Object Counting System (ISOCS) software was used to model efficiency curves for each location measured with the high-purity germanium (HPGe) detector. The geometry templates and specific ISOCS inputs for measurement locations in the spent fuel pool (SFP)/transfer canal are discussed below. ORISE has previously reviewed ZSs technical support document (TSD) for performing in situ gamma spectrometry measurements and concluded that the modeling approach was representative of site-specific conditions (ORISE 2015). Therefore, ORISE used the measurement templates and inputs, as deemed appropriate, for the measurement locations presented herein.

A circular plane geometry was used to assess radionuclide of concern (ROC) concentrations in measurements collected from this area. An illustration of the circular plane geometry appears in Figure C.1. The geometry template was modeled in the same manner as presented in ZSs TSD. The circular plane model consists of a concrete source layer of thickness of 0.0508 meter (m)

(corresponding to a 2-inch depth) covering a non-radioactive concrete backing of 0.3 m. Source-to-

Zion Spent Fuel Pool and Transfer Canal C-3 5271-SR-05-0 detector distance varied from one measurement location to another to be consistent with the standoff parameters used by ZS at these locations, and this value ranged from 1.03 m to 3.05 m.

Figure C.1. ISOCS Circular Plane Template (Canberra 2009)

C.4 RADIOLOGICAL SAMPLE ANALYSIS C.4.1 Gamma Spectroscopy Samples were analyzed as received and mixed, crushed, and/or homogenized, as necessary, and a portion sealed in a 0.06-liter Marinelli beaker. The quantity placed in the beaker was chosen to reproduce the calibrated counting geometry. Net material weights were determined and the samples counted using intrinsic, HPGe detectors coupled to a pulse-height analyzer system. Background and Compton stripping, peak search, peak identification, and concentration calculations were performed using the computer capabilities inherent in the analyzer system. All total absorption peaks (TAPs) associated with the ROCs were reviewed for consistency of activity. Spectra also were reviewed for other identifiable TAPs. TAPs used for determining the activities of radionuclides and the typical associated minimum detectable concentrations (MDCs) for a 4-hour count time are presented in Table C.1.

Table C.1. Typical MDCs Total Absorption Peaks Radionuclide TAP (MeV)a MDC (pCi/g)

Co-60 1.332 0.09 Cs-134 0.795 0.09 Cs-137 0.662 0.07 aMeV = mega electron volt

Zion Spent Fuel Pool and Transfer Canal C-4 5271-SR-05-0 C.4.2 Ni-63 Analysis Samples were spiked with a nickel and cobalt carrier and digested with a mixture of nitric and hydrochloric acids. Unwanted elements, such as iron and cobalt, are then removed by running the slurry via anion exchange chromatography. Nickel is then separated from the slurry using a nickel selective resin cartridge. The purified nickel is then eluted off of the column with a dilute nitric acid solution. Ni-63 activity is then determined via liquid scintillation counting. The typical MDC for a 2-gram sample and 60-minute count time using this procedure is 0.6 pCi/g.

C.4.3 Radioactive Strontium Analysis Sr-90 concentrations were quantified by total sample dissolution followed by radiochemical separation and counted on a low background proportional counter. Samples were homogenized and dissolved by a combination of potassium hydrogen fluoride and pyrosulfate fusions. The fusion cakes were dissolved, and strontium was co-precipitated on lead sulfate. The sulfate-salt complex was dissolved in EDTA at a pH of 8.0. The strontium was separated from residual calcium and lead by reprecipitating strontium sulfate from EDTA at a pH of 4.0. Strontium was separated from barium by complexing the strontium in DTPA while precipitating barium as barium chromate. The strontium was ultimately converted to strontium carbonate and counted on a low-background gas proportional counter. The typical MDC for a 60-minute count time using this procedure is 0.4-0.6 pCi/g depending on the Y-90 ingrowth at time of counting.

C.4.4 H-3 Analysis Tritium (H-3) analyses were performed using a material oxidizer and counted by liquid scintillation.

The Material Oxidizer combusts samples in a stream of oxygen gas and passes the products (including CO2 and H2O vapor) through a series of catalysts. The H-3 is carried by water and is captured in a trapping scintillation cocktail specific to water. The typical MDC for H-3 for a 60-minute count time using this procedure is 3-5 pCi/g.

Zion Spent Fuel Pool and Transfer Canal 5271-SR-05-0 APPENDIX D: MAJOR INSTRUMENTATION

Zion Spent Fuel Pool and Transfer Canal D-1 5271-SR-05-0 The display of a specific product is not to be construed as an endorsement of the product or its manufacturer by the author or his employer.

D.1 SCANNING AND MEASUREMENT INSTRUMENT/DETECTOR COMBINATIONS D.1.1 Gamma Ludlum NaI Scintillation Detector Model 44-10, Crystal: 5.1 cm x 5.1 cm Coupled to: Ludlum Ratemeter-scaler Model 2221 Coupled to: Trimble Geo 7X High Purity, Broad-Energy Germanium Detector Canberra Model No. BE3825 Used in conjunction with:

Canberra Inspector 2000 multi-channel analyzer, Canberra In-Situ Object Counting System and Genie 2000 software, Canberra 50 mm, 90-degree FOV lead collimator, and Dell laptop (Canberra, Meriden, Connecticut)

D.2 LABORATORY ANALYTICAL INSTRUMENTATION High-Purity, Extended-Range Intrinsic Detector Canberra/Tennelec Model No: ERVDS30-25195 Canberra Lynx Multichannel Analyzer Canberra Gamma-Apex Software (Canberra, Meriden, Connecticut)

Used in conjunction with:

Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, Tennessee) and Dell Workstation (Canberra, Meriden, Connecticut)

High-Purity, Intrinsic Detector EG&G ORTEC Model No. GMX-45200-5 Canberra Lynx Multichannel Analyzer Canberra Gamma-Apex Software (Canberra, Meriden, Connecticut)

Used in conjunction with:

Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, Tennessee) and Dell Workstation (Canberra, Meriden, Connecticut)

High-Purity, Intrinsic Detector EG&G ORTEC Model No. GMX-30P4 Canberra Lynx Multichannel Analyzer Canberra Gamma-Apex Software

Zion Spent Fuel Pool and Transfer Canal D-2 5271-SR-05-0 (Canberra, Meriden, Connecticut)

Used in conjunction with:

Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, Tennessee) and Dell Workstation (Canberra, Meriden, Connecticut)

High-Purity, Intrinsic Detector EG&G ORTEC Model No. CDG-SV-76/GEM-MX5970-S Canberra Lynx Multichannel Analyzer Canberra Gamma-Apex Software (Canberra, Meriden, Connecticut)

Used in conjunction with:

Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, Tennessee) and Dell Workstation (Canberra, Meriden, Connecticut)

Low-Background Gas Proportional Counter Series 5 XLB (Canberra, Meriden, CT)

Used in conjunction with:

Eclipse Software Dell Workstation (Canberra, Meriden,CT)

Liquid Scintillation Analyzer Perkin Elmer Model Tri-Carb 5100 TR (Perkin Elmer, Shelton, CT)

Used in conjunction with:

Quantamart Software Perkin Elmer, Shelton, CT)

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