Text

Independent Confirmatory Survey Summary and Results for the Containment and Auxiliary Buildings at the Zion Nuclear Power Station, Zion, Illinois
	ML18285A143
Person / Time
Site:	Zion File:ZionSolutions icon.png
Issue date:	10/09/2018
From:	Altic N; Oak Ridge Institute for Science & Education
To:	John Hickman; Reactor Decommissioning Branch
References
	Contract No. DE-SC0014664, DCN 5271-SR-03-0, RFTA NO.18-005
	Download: ML18285A143 (85)
	v • d • e

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OAK RI DGE INSTITUTE FOR SCIENCE AND EDUCATION Shaping t he Future of Science October 9, 2018 Mr. John Hickman U.S. N uclear Regulatory Commission O ffice of Nuclear Material Safety and Safeguards Division of D ecommissioning, Uranium Recovery, and Waste Programs Reactor D ecommissioning Branch Mail Stop: T8F5 11545 Rockville Pike Rockville, MD 20852

SUBJECT:

DOE Contract No. DE-SC0014664 INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE CONTAINMENT AND AUXILIARY BUILDINGS AT THE ZION NUCLEAR POWER STATION, ZION, ILLINOIS (RFTA NO.18-005); DCN 5271-SR-03-0

Dear Mr. Hickman:

The O ak Ridge Institute for Science and Education (ORISE) is pleased to provide the enclosed report detailing the independent confirmatory survey activities associated with the Containment and Auxiliary Buildings at the Zion N uclear Power Station in Zion, Illinois. This report provides the summary and results of ORISE on-site activities performed during the period of April 16-26, 2018.

NRC comments on the draft report have been incorporated into this final version.

You may contact me at 865.574.6273 or Erika Bailey at 865.576.6659 if you have any questions.

Nick A. Altic, CHP Health Physicist/Project Manager O RISE NAA:j c electronic distribution: B. Lin,NRC R. Edwards, N RC T. Carter, NRC M. Kunowski, NRC K. Conway, NRC E . Bailey, ORISE S. Roberts, ORISE D . Hagemeyer, ORISE File/5271 100 ORAU Way* Oak Ridge* TN 37830

orise.orau.gov

OAK RIDGE INSTITUTE FOR SCIENCE ANO EDUCATION Shaping the Fu ture of Scienc e INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE CONTAINMENT AND AUXILIARY BUILDINGS AT THE ZION NUCLEAR POWER STATION ZION, ILLINOIS N. A. Altic, CHP ORISE FINAL REPORT Prepared for the U .S. Nuclear Regulatory Commission October 2018 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 a

teams of experts, unique laboratory capabilities and access to consortium of more than 100 major Ph.D.-granting institutions, ORAU works with federal, state, local and commercial customers to advance national priorities andrserve 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 State~ 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.

INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR

)

THE CONTAINMENT AND AUXILIARY BUILDINGS AT THE ZION NUCLEAR POWER STATION, ZION, ILLINOIS

OAK RIDGE INSTITUTE

'~ I FOR SCI,ENCE AND EDUCATION .

)

Prepared by N. A. Altic, CHP ORISE FINAL Prepared for the U.S. Nuclear Regulatory Commission OCTOBER 2018 This doi:uinent was prepared for the U.S. Nuclear Regulatory Commission by the Oak Ridge Institute for Science and Education (ORISE) through an interagency agreement with the U.S.

Department of Energy (DOE) (NRC FIN No. F-1244). ORISE is managed by Oak Ridge Associated Universities under DOE contract number DE-SC0014664.

Zion Containment and Auxiliary Buildings 5271-SR-03-0

OAK RIDGE INSTITUTE I* I FOR SCIENCE AND EDUCATION INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE CONTAINMENT AND AUXILIARY BUILDINGS AT THE ZION NUCLEAR POWER STATION, ZION, ILLINOIS Prepared by: Date:

N. A. Altic, CHP, Health Physicist/Project Manager ORISE Reviewed by: Date: / ~/9)21) <f .

P. H. Benton, Quality Manager ORISE Date:

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Reviewed by:

Date:

Reviewed by:

W. F. Smith, Senior Chemist .

ORISE Reviewed and approved for release by: ~ .

'E. N. Bailey, ORISE FINAL REPORT October 2018 Zion Containment and Auxiliary Buildings 5271-SR-03-0

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~ ~ FOR SCIENCE AND EDUCATION CONTENTS FIGURES ................ ,....................................................... '. ........................................................................... :..... iii TABLES ......................................................................................................................................................... .iv ACRONYMS ..................................................................................................................................................... v EXECUTIVE

SUMMARY

............................................................................................................................vii

1. INTRODUCTION ..............................................................................................................*......................... 1

2. SITE DESCRIPTION ................................................................................................................................. 2 2.1 Containment Buildings ................................................................................. ,................................. 3 2.2 Auxiliary Building ......................................*............. :........................................................................ 3

3. DATA QUALITY OBJECTIVES ............................................................................................................. 4 3.1 State the Problem ................................................................................................. :. ......................... 5 3.2 Identify the Decision ...................................................................................................................... 5 3.3 Identify Inputs to the Decision ..................................................................................................... 6 3.3.1 Radionuclides of. Concern and Release Guidelines .............................................................. 7 3.4 Define the Study Boundaries., ....................................................................................................... 9 3.5 Develop a Decision Rule ............................................................................................................. 10 3.6 Specify Limits on Decision Errors ............................................................................................. 10 3.7 Optimize the Design for Obtaining Data ............................................................................... :.. 11 3.7.1 SOF Calculation for the Containment Buildings ............................................................... 11

4. PROCEDURES *......................................................................................................................................... 11 4.1 Reference Syste_m .................... ,. .................................. :.................................................................. 12 4.2 Surface Scans .......................................... .-...................................................................................... 12 4.3 Measurement/Sampling Location Determination .............................................. ,.................... 12 4.4

ISOCS Measurements ................................... '. ............................ ,................................................. 14 4.5 Volumetric Samples ...................................................................................................................... 15

5. SAMPLE ANALYSIS AND DATA INTERPRETATION ............................................................... 15

6. FINDINGS AND RESULTS ................................................................................................................... 16 6.1 Unit 1 Containment .............. :..............................'......................................................................... 16 6.1.1 Unit 1 Undervessel Area ........................................................................................................ 16 6.1.2 Unit 1 Containment 565-ft Elevation .................................................................................. 21 6.2 Unit 2 Containment ............................................... :...................................................................... 25 6.2: 1 Unit 2 Undervessel Area .............................................. ,......................................................... 25 6.2.2 Unit 2 Containment 565-ft Elevation ............................................................... :.................. 30 6.3 Auxiliary Building......................................................................................................................... 33 Zion Containment and Auxiliary Buildings 5271-SR-03-0

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I . I I FOR SC[ENCE AND EDUCATlON 6.3.1 Surface Scans ............................................................................................................................ 33 I

6.3.2 ROC Concentrations in Volumetric Samples ..................................................................... 35 6.3.3 Auxiliary Building Floor ROC Concentration Assessment .............................................. 38

7.

SUMMARY

AND CONCLUSIONS ...................................................................................................... 40 7.1 Containment Buildings ................................................................................................................. 40 7.2 Auxiliary Building ..................................................... :.................................................................... 41

8. REFERENCES ........................................................................................................................................... 43 APPENDIX A: FIGURES APPENDIX B: DATA TABLES APPENDIX C: SURVEY AND ANALYTICAL PROCEDURES APPENDIX D: :MAJOR INSTRUMENTATION Zion Containment and Auxiliary Buildings 11 5271-SR-03-0

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~ ~ FOR SCIENCE AND EDUCATION FIGURES Figure 2.1. ZNPS Overview (adapted from ZS 2017) ...................................... ,.......................................... 2 Figure 2.2. ZNPS Containment and Auxiliary Buildings (Google Earth) ................................................ .4 Figure 4.1. Sample Size Determination Using VSP .................................................................................... 14 Figure 6.1. Q-plot for Gamma Scan Data of Unit 1 U ndervessel Area Floor and Lower Walls ......... 16 Figure 6.2. Comparison of PSS Data and ORISE Confirmatory Mean Concentrations and Uncertainties for Gamma-emitting Radionuclides in the Unit 1 Undervessel Area ... :....... 18 Figure 6.3. Q-Q Plot for Gamma Scan Data of Unit 1 565-ft Elevation ................................................ 22 Figure 6.4. Example Location of Elevated Direct Gamma Radiation Identified on Lower Walls of the Unit 1 565-ft Elevation ............................................................................................... :.......... 23 Figure 6.5. Comparison of PSS Data and ORISE Confirmatory Mean Concentrations and Uncertainties for Gamma-emitting Radionuclides in the Unit 1 565-ft Elevation ............. 25.

Figure 6.6. Q-plot for Gamma Scan Data of Unit 2 Undervessel Area Floor and Lower Walls ......... 26 figure 6.7. Comparison of PSS Data and ORISE Confirmatory Mean Concentrations and Uncertainties for Gamma-emitting Radionuclides in the Unit 2 Undervessel Area ........... 28 .

Figure 6.8. Q-Q Plot for Gamma Scan Data of Unit 2 565-ft Elevation ................................................ 31 Figure 6.9. Comparison of PSS Data and ORISE Confirmatory Mean Concentrations and Uncertainties for Gamma-emitting Radionuclides in the Unit 2 565-ft Elevation ............. 33 Figure 6.10. Q-plot for the Auxiliary Building Floor and Sumps ............................................................. 34 Figure 6.11. Q-plot for the Auxiliary Building West Wall ......................................................................... 35 Figure 6.12. Hot Particle Separated from Concrete Sample M0020 ........................................................ 36 Figure 6.13. Comparison of PSS Data and ORISE Confirmatory Mean Concentrations and Uncertainties for Gamma-emitting Radionuclides in the Auxiliary Building Floor ............ 39 Zion Containment and Auxiliary Buildings iii 5271-SR-03-0

TABLES Table 3.1. ZNPS Confirmatory Survey Decision Process .......................... :................................................ 6 Table 3.2. ZNPS Basement Surfaces DCGLs* .............................................................................................. 8 Table 3.3. PSS SU Designations ...........................................*........................................................................... 9 Table 6.1. Summary of Unit 1 Undervessel Confirmatory In-Situ Gamma Spectrometry Measurements ................................................................................................................................ 17 Table 6.2. Summary of Unit 1 Undervessel Random Volumetric Concrete Samples ........................... 19 Table 6.3. Unit 1 Undervessel Results by Analysis Method ...................................................................... 20 Table 6.4. Summary of Unit 1 565 ft Elevation Confirmatory In-Situ Gamma Spectrometry Measurements ........................................................................ ,............. :......................................... 24 Table 6.5. Summary of Unit 2 Undervessel Confirm:i.tory In-Situ Gamma Spectrometry Measurements ................................................................................................................................ 27 Table 6.6. Summary of Unit 2 Undervessel Random Volumetric Concrete Samples ........................... 29 Table 6.7. Unit 2 Undervesssel Results by Analysis Method ............................................................... ,..... 30 Table 6.8. Summary of Unit 2 565-ft Elevation Confirmatory In-Situ Gamma Spectrometry Measurements ................................................................................................................................ 32 Table 6.9. Summary of ROC Con:ce~trations in the Auxiliary Building Floor .................................. :.... 35 Table 6.10. Summary of HTD ROC Concentrations in Select Auxiliary Building Concrete Samples37 T:i,ble 6.11. ROC Concentrations in Water/Sediment Auxiliary Building Sump Samples .................... 38 Table 6.12. Assessment of ROC Concentrations in the Auxiliary Building Floor ................................ .40 Zion Containment and Auxiliary Buildings 1V 5271-SR-03-0

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~ f2 FOR SCIENCE AND EDUCATION ACRONYMS AA alternate action cpm counts per minute DCGL derived concentration guideline level DCGL~c Base Case DCGL DCGL0p OperationalDCGL DOE U.S. Department of Energy DQO data quality objective DS decision statement ECP exponential circular plane EPA U.S. Environmental Protection Agency Exelon Exelon Generation Company FOV field of view FSS final status survey ft feet HPGe high-purity germanium HID hard-to-detect HUT hold-up tank ISOCS In Situ Object Counting System LACE Line Activity Consistency Evaluator LTP license termination plan l\1ARSSIM Multi-Agency Radiation Suzyey and Site Investigation Manual MDC minimum detectable concentration mrem/yr millirem per year Nal sodium iodide NIST National Institute of Standards and Technology NRC U.S. Nuclear Regulatory Commission ORISE Oak Ridge Institute for Science and Education pCi/g p1cocunes per gram pCi/m2 picocuries per square meter PSQ principal study question Q-plot quantile plot Q-Qplot quantile-quantile plot REAL Radiological and Environmental Analytical Laboratory RL relaxation length ROC radionuclide of concern RPV reactor pressure vessel RSS ranked set sampling SFP spent fuel pool SOF sum-of-fractions SU survey unit TAP total absorption peak TEDE total effective dose equivalent TSD technical support document Zion Containment and Auxiliary Buildings V 5271-SR-03-0

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Si *~ FOR SCIENCE AND EDUCATION UCL upper confidence level VSP Visual Sample Plan WWTF Waste Water Treatment Facility ZNPS Zion Nuclear Power Station ZS ZionSo!utions, LLC Zion Containment and Auxiliary Buildings vi 5271~SR-03-0

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~ ~ FOR SCIENCE AND EDUCATION INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR

.THE CONTAINMENT AND AUXILIARY BUILDINGS 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 Containment Buildings and Auxiliary Building at the Zion Nuclear Power Station.

Confirmatory survey activities were conducted during the period of April 16-26, 2018 and consisted of: surface scans, in sitt1 gamma spectrometry measurements, volumetric concrete sampling, and water and sediment sampling.

For gamma-emitting radionuclides of concern in the Containment Buildings-Cs-134, Cs-137, Co-60, Eu-152, and Eu-154--0RISE did not identify issues that contradicted the final status survey (FSS) data for demonstrating compliance with the release criterion. The hard-to-detect (HID) concentrations for Sr-90 and Ni-63 in the Containment Building were unremarkable. The tritium concentration fraction in the Undervessel Area for both Containment Buildings exceeded 1.00.

Additionally, three individual measurements in the Unit 1 Undervessel Area and one individual measurement in the Unit 2 Undervessel Area exceeded the Base Case derived concentration guideline level (DCGL8 c) for tritium. The tritium concentration was based on volumetric samples that represented a depth of six inches. ORISE recommends that NRC evaluate the potential for tritium contamination greater than six inches in the Undervessel Area concrete.

Based on the volumetric confirmatory survey samples, the FSS in situ gamma spectrometry

.i measurements in the Auxiliary Building provided a conservative representation of the residual radioactivity. ORISE collected two concrete samples from the west wall (northern most Hold-up Tank [HUT] Cubicle) and a floor location where the Cs-137 concentrations exceeded the DCGLiic*

These two locations were from areas with localized elevated direct gamma radiation. The ISOCS FSS measurements were less than the DCGL0p; the difference due to the much larger area represented by the gamma spectrometry measurement relative to the concrete sample. The resulting upper bound of confirmatory gamma-emitting ROC SOF for the Auxiliary Building Floor, including hot spots, was a small fraction relative to the DCGLs. The HID results for the Auxiliary Building Zion Containment and Auxiliary Buildings vii 5271-SR-03-0

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~ ~ FOR SCIENCE AND EDUCATION did not indicate that the approved ratios presented in the Hcense termination plan were non-conservative.

Confirmatory survey results from sediment and water sampling indicated that there is mobile radioactivity present in the Auxiliary Building Basement. At this time, ORISE cannot confirm the source of the contaminated particulates present in the sump from where these samples were collected.

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~ i FOR SCIENCE AND EDUCATION INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE CONTAINMENT AND AUXILIARY BUILDINGS 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 pools. Both units were then placed in safe storage pending the commencement of site decommissioning and dismantlement. In 2010, the 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 ten 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 unr,estricted use. Upon successful completion of the decommissioning activities, control and responsibility for the site will be transferred back to Exelon and the independent spe~t fuel ~torage installation maintained under Exelon's Part SO license (EC 2015).

As part of decommissioning, all above-grade structures, with minor exceptions, will be demolished.

Structures below the 588-ft elevation (referenced from mean sea level), consisting of primarily exterior sub-grade walls and floors, will remain. These basement.structures will be backfilled as part of the final site restoration. In order to demonstrate compliance with the release criteria in 10 CFR 20.1402, ZS will implement a FSS 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 ZS's license termination plan (LTP) (ZS 2017). The primary FSS method for basement structure survey units (SUs) is in situ gamma measurements using a portable, high-resolution gamma spectrometer. FSS methods are based on methods outlined in the Multi-Agenry Radiation Survry and Site Investigation Manual (MARSSIM) (NRC 2000).

Zion Containment and Auxiliary Buildings 1 5271-SR-03-0

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..., ~ FOR SC l ENCE AND EDUCATlON The N RC requested ORISE to perform confirmatory survey activities of the remaining structures and surfaces of the Containment and Auxiliary Buildings at the ZNPS. This report summarizes the confirmatory survey activities and results for these areas.

2. SITE D ESCRIPT ION The 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 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 2017). 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 ft above mean sea level; for reference, the elevation of Lake Michigan-which bounds the site on the east-is approximately 577.4 ft at low water level (ZS 2017).

Legend Part 50 Uc nse

- Boundary Radiologically

" - " Restricted Area Fence Security Restricted ~~H"~ ~ !tt.r

- **- Area Fence *~

Figure 2.1. ZNPS Overview (adap ted from ZS 2017)

Zion Containment and Auxiliary Buildings 2 5271-SR-03-0

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..., ~ FOR SCIENCE AND EDUCATION 2.1 CONTAINMENT BUILDINGS T he Containment Buildings housed the reactor pressure vessels, and each building consists of a steel liner with interior and exterior concrete several feet thick. In both Containment Building basements, all concrete from the interior of the steel liner from the 565-ft to the 588-ft elevation was removed.

The basement structure below the 565-ft elevation is referred to as the Undervessel Area and contains the In-core Instrument Shaft. Approximately 15 centimeters of concrete has been removed as part of the remediation process in the Unit 1 Undervessel Area. The remediation extent was more extensive in the U nit 2 Undervessel Area than in Unit 1.

2.2 AUXILIARY BUILDING The Auxiliary Building had the various support systems for both reactors, such as residual heat removal and letdown system. The Auxiliary Building basement consists of wall and floor structures below the 588-ft elevation. There are two sumps located in the Auxiliary Building: Sump Pit A and Sump Pit B (see Figure A-5). The bottom of the two sumps are located at slightly different elevations 533 ft for Sump Pit A and 536-ft for Sump Pit B. Also present is the elevator pit, the bottom of which is located on the 536-ft elevation. Auxiliary Building structures above the 588-ft elevation have been demolished, leaving the basement open to the environment.

Figure 2.2 displays the licensed area with the Auxiliary and Containment Buildings indicated.

Zion Containment and Auxiliary Buildings 3 5271-SR-03-0

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~ ~ FOR SCIENCE AND EDUCATION Figure 2.2. ZNPS Containment and Auxiliary Buildings (Google Earth)

3. DATA QUALITY OBJECTIVES The data quality objectives (DQOs) described herein are consistent with the Guidance on Systematic Planning Using the Data Quality O~jectives 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 Zion Containment and .Auxiliary Buildings 4 5271 -SR-03-0

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~ ~ FOR SCIENCE AND EDUCATION

4. Define the study boundaries

5. Develop a decision rule

6. Specify limits on decision errors

7. Optimize the design for obtaining data 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 NRC requested that ORISE perform confirmatory surveys at the ZNPS. The objective of the confirmatory surveys was to provide NRC with independent confirmatory data for NRC's consideration in_ the evaluation of the PSS results. The problem statement was formulated as follows:

Confirmatory surveys are necessary to generate independent radiological data for NRC's consideration in the evaluation of the PSS design, implementation, and results for demo~strating compliance with the release criteria.

3.2 IDENTIFY THE DECISION The second step in the DQO process identified the principal study questions (PSQs) and alternate actions (AAs), developed a decision statement, 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 decision statement. The PSQ, AAs, and combined decision statement (DS) are organized and presented in Table 3.1.

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~ ~ FOR SCIENCE AND EDUCATION Table 3.1. ZNPS Confirmatory Survey Decision Process Principal Study Questioq Alternate Actions Yes:

Compile confirmatory data and report results to the 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 Do confirmatory survey results agree with the statistical sample population examination/ assessment final radiological survey data for the Unit 1 and conditions were met.

Unit 2 Containment Buildings and the Auxiliary No:

Building?

Compile confirmatory data and report results to the NRC for their decision making. Provide independent interpretation of confirmatory survey results identifying any anoq-ialous field or laboratory data and/ or when statistical sample population examination/ assessment conditions were not satisfied for the NRC's determination of the adequacy of the FSS data.

Decision S~atement Confirmatory survey result~ 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 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 met data requirements. For this effort, information inputs included the following:

Concrete characterization data for the Auxiliary and Unit 1 and Unit 2 Containment Buildings (NRC provided ORISE the Excel files containing the characterization data for these areas.)

FSS data for the Auxiliary and Unit 1 and Unit 2 Containment Buildings (NRC provided ORISE the data analysis Excel files for these areas.)

Derived concentration guideline levels (DCGLs), discussed in subsection 3.3.1

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 sample analysis results for concrete samples Zion Containment and Auxiliary Buildings 6 5271-SR-03-0

OAK RIDGE INSTITUTE I I FOR SCIENCE AND EDUCATION 3.3.1 Radionuclides of Concern and Release Guidelines The primary radiomiclides identified for the ZNPS are beta-gamma emitters-fission and.activatio_n products-resulting from reactor operations. At ZNPS, there are four distinct source terms:

basement structures, soils, buried piping, and groundwater. Furthermore, basement structures comprise four structural source terms: surfaces, embedded piping, penetrations, and fill. ZS*has 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 DCGLs-defined in ZS's LTP as Base Case DCGLs (DCG~cs)-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's release criteria, the DCGLiics are further reduced to Operational DCGLs (DCGL0ps). The DCGL0ps are scaled to an expected dose from prior investigations and are used for remediation and FSS design purposes. The initial suite of radionuclides present at ZNPS has been reduced based on an insignificant dose contribution from a number of radionuclides. The DCGLBcs and DCGL0ps, accounting for insignificant dose contributors, for the basement structure source terms-excluding fill material-are presented in Table 3.2.

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IS§ ~ FOR SClENCE AND EDUCATION Table 3.2. ZNPS Basement Surfaces DCGLsa Containment Turbine Building Circ Crib*

Auxiliary SFP /Transfer Floors ROC Above

Under~ Water House* :.ww'i'F.

Building Canal & *"

. 565 ft vessel Discharge /Fotebay Wallsh Tunnel '.

Base Case DCGLs (pCi/m2)

H-3 5.30E+08 2.38E+08 2.38E+08 l.29E+08 l.93E+08 l.71E+07 Co-60 3.04E+08 l.57E+08 l.57E+08 7.03E+07 5.52E+07 2.83E+07 Ni-63 l.15E+l0 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 l.59E+07 2.13E+07 2.31E+06 Cs-137 l.11E+08 3.94E+07 3.94E+07 2.11E+07 2.96E+07 2.93E+06 Sr-90 9.98E+06 l.43E+06 l.43E+06 7.74E+05 l.16E+06 l.03E+05 Eu-6.47E+08 3.66E+08 3.66E+08 l.62E+08 l.23E+08 7.55E+07 152 Eu-5.83E+08 3.19E+08 3.19E+08 l.43E+08 l.12E+08 5.74E+07 154 Operational DCGLs (pCi/m2) . . ,. ..

H-3 l.71E+08 3.25E+07 2.37E+08 4.98E+07 l.10E+07 5.39E+07 7.43E+07 3.28E+06 Co-60 9.81E+07 2.15E+07 l.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 l.85E+08 9.11E+08 l.25E+09 5.55E+07 Cs-134 6.81E+07 4.12E+06 2.99E+07 6.30E+06 l.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 l.79E+O~ 8.82E+06 L14E+07 5.63E+05 Sr-90 3.22E+06 l.96E+05 1.42E+06 2.99E+05 6.58E+04 3.24E+05 4.47E+05 l.98E+04 Eu-2.09E+08 5.00E+07 3.64E+08 7.66E+07 l.38E+07 6.77E+07 4.74E+Ci7 l.45E+07 152 Eu-

.i l.88E+08 4.36E+07 3.17E+08 6.67E+07 l.22E+07 5.98E+07 4.31E+07 l.10E+07 154

Recreated from ZS 2017 hThe Operational DCGLs 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 DCG:Lncs 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 are performed as follows:

SOF = Ln .

]=1 Cmean,j DCGLsc,j

+ (CELv,j-Cmean,j)

(ncGL BC,J*x(SAsu SAElv.

))

Eq. 3-1 Zion Containment and Auxiliary Buildings 8 5271-SR-03-0

OAK RIDGE INSTITUTE I* ~ FOR SCIENCE AND EDUCATION Where cmean)s the mean concentration of ROC "j," CElv,jis an elevated area of ROC "j," DCGLnc,jis the Base Case DCGL for ROC "j," SAsu is the adjusted surface area of the PSS unit, and SAEiv. is the surface area of the elevated measurement. Per Section 5.5.4 of the LTP, areas of elevated activity-for building surfaces-are defined as any area identified by measurement/ sample (systematic or judgmental) that exceeds the Operational DCGL but is less than the Base Case DCGL. Any area that exceeds the Base Case DCGL will be remediated. 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 in both the Unit 1 and Unit 2 Containment Buildings and the floor of the Auxiliary Building. For the purpose of this confirmatory survey, two separate building surface survey units were considered for each containment building as follows:

1) the Undervessel Area and 2) the 565-ft Elevation and Above. The PSS SU IDs for the subject confirmatory survey areas are summarized in Table 3.3.

Table 3.3. PSS SU Designations

- Area SUID Unit 1 Undervessel Area B101110AF Unit 1 565-ft Elevation and Above B101100AF Unit 2 Undervessel Area B102110AF Unit 2 565-ft Elevation and Above B102100AF Auxiliary Building B105100AF Confirmatory surveys in the Auxiliary Building basement focused on the remaining concrete floor.

At NRC's request, confirmatory survey measurements were collected from the Auxiliary Building west wall near the former north Hold-up Tank (HUT) cubicle. Sumps were investigated as part of the SU s in which they reside. The confirmatory survey activities were conducted during the period

of April 16-26, 2018.

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~ ~ FOR SCIENCE AND EDUCATION 3.5 DEVELOP A DECISION RULE The fifth step in the DQO process specified appropriate population parameters (e.g., mean, median), confirmed action levels were above detection limits, and developed an if... then ... decision rule statement: Decision rules for this survey ,are based on independent scan surveys, in situ gamma spectroscopy measurements, and concrete sample results to assess whether there is a statistical bias rel~tive to the FSS data. Typically, decision rules are based on a statistical comparison of the ORJSE survey data and the FSS data using an appropriate test. However, the difference in size between sample populations was significant. For example, 60 FSS in situ gamma spectroscopy measurements were collected in the Unit 1 Undervessel Area SU, whereas ORJSE collected 9 measurements. The approximately 6:1 sample size ratio reduces the power to detect smaller but potentially meaningful statistical difference between the two sample groups. Therefore, alternative assessment methods were employed.

Different sample types were collected from the Containment Buildings and Auxiliary Building (i.e., in situ gamma spectroscopy measurements in the Containment Buildings and concrete samples in the Auxiliary Building). The parameter of interest for each survey area is the mean ROC concentration/inventory and the associated 95% confidence interval of the mean.

While the sample media are different, the parameter of interest for this study is the same for each area. The aforementioned information is combined to formulate the decision rule for the Containment and Auxiliary Buildings-which is stated as follows:

If the mean ROC concentrations/ inventory of the confirmatory and FSS sample populations overlap at the 95% confidence level and results arc below the applicable limit, conclude that confirmatory surory data agrees with the FSS dat~othernnse, pciform fi1rther cvaluation(s) andprovide technical comments to the 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 of decision errors. Decision errors were controlled both during the confirmatory survey investigations and during data quality assessment and were based on two orders of control.

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i I

OAK RIDGE INSTITUTE I I FOR SCIENCE AND EDUCATION 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 depended on the FSS estimated mean not being larger than the difference between the DCGL and the FSS estimated mean (i.e. 1.00 - Mean SOF). The confirmatory survey mean was estimated at the 95% confidence level.

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

Measurement MDCs for the in situ gamma spectrometry measurements were less than 50% of the I 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.

3. 7 OPTIMIZE THE DESIGN FOR OBTAINING DATA The seventh step in the DQO process is 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.

3.7.1 SOP Calculation for the Containment Buildings Physical boundaries and access constraints in both Unit 1 and Unit 2 Containment Buildings limited ORISE from achieving 100% gamma scan coverage. Thus, all of the potential hot spots were not identified and considered in the mean concentration. Therefore, the 95-percent upper confidence level (UCL95) of the mean was used for the SOF calculation. The UCL accounts for the uncertainty in the estimate of the mean due to sampling error, the probability of a Type I decision error is minimized (i.e., incorrectly concluding that the ROC concentration is less than DCGl.iJc).

4. PROCEDURES The ORISE survey team conducted independent confirmatory survey activities, including surface scans, in situ gamma spectrometry measurements, and volumetric sampling activities, within those Zion Containment and Auxiliary Buildings 11 5271-SR-03-0

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~ ~ FOR SCIENCE AND EDUCATION accessible survey areas specifically requested by the NRC. Survey activities were conducted in accordance with the ORAU Radiological and Environmental Survry 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. Specific areas were also digitally photographed.

4.2 SURFACE SCANS Thallium doped sodium iodide (NaI[I1J) detectors were used to evaluate direct gamma radiation levels on basement surfaces. The Auxiliary Building basement floor received high-density scan coverage. The northern portion of the west wall of the Auxiliary Building received high density scans on the lower portion and medium-density scans on the upper portion, near the former northern most HUT cubicle. High-density gamma scans were performed on the lower walls and floor of the Containment Undervessel Area and 565-ft Elevation and Above.

All detectors were coupled to Ludlum Model 2221 ratemeter-scalers with audible indicators *and were also coupled to data-loggers to electronically record all scanning data for surface structures.

Locations of elevated response that were audibly distinguishable from local background levels, suggesting the presence of elevated 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. The planned measurement type depended on the SU in which the measurement was to be collected-for example, concrete samples in the Auxiliary Building basement and in situ gamma spectroscopy measurements for the Containment Buildings SUs. For the Auxiliary Building floor, a ranked set sampling (RSS) process, following U.S. Environmental Protection* Agency (EPA) guidance, was used to designate random locations (EPA 2002).

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OAK RIDGE INSTITUTE I I FOR SCIENCE AND EDUCATION RSS provides a methodology to determine the necessary number of samples to estimate the mean concentration of a population; however, it does not require the assumption of a normal distribution~

The process combines random sampling with the use of a field screening method capable of distinguishing the relative magnitude of a parameter of interest in a population combined with professional judgment to select sampling locations. Field screening was accomplished with portable Nal gamma detectors.

The RSS systematic planning process uses a replication method on a larger random population from which the locations for the resulting samples were selected. Replication refers to the number of cycles (r) for performing a set size (m) of field measurements. The set size was maintained at three locations (m = 3) to minimize ranking errors. The number of assessment locations per cycle is dependent on the set size and is simply m2

Therefore, in a given cycle, samples are collected from each set based on the following ranking criteria:

Set 1: The lowest of three gamma measurement locations within Set 1 is sampled.

Set 2: The middle of three gamma measurement locations within Set 2 is sampled.

Set 3: The highest ofthree gamma-measurement locations within Set 3 is sampled.

The number of repetitive cycles was dependent on the total number of samples/measurements (n) required and is a function of n and m-simply defined as n =m x r. For the Auxiliary Blililding floor, the PSS in sitt1 gamma spectroscopy data indicated a mean SOP of 0.08 and a corresponding standard deviation of 0.11. Using these values as inputs, VSP calculated that 9 samples were needed (n =m x r; 9 =3 x 3), collected from 27 field ranking locations, to conclude that the estimated mean SOP falls between 0.01 and 0.15 at the 95% confidence level. Figure 4.1 illustrates the sample size determination using VSP.

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~ ~ FOR SClENCE AND EDUCATION

~ How Many Samples ke Needed'>

~ Design Parameters A balanced desigi will be used for the symmetric distribution.

I want to be 195.00 % confident that the estimated mean is within lo.07 units jabove or below ..:.J the true mean. (Two-sided confidence interval)

I estimate the standard deviation to be 10.11 r How Many Samples If Simple Random Sampling Were Used?

For simple random sampling, 12 samples would be needed.

~ How Many Samples Needed For Ranked Set Sampling?

Chosen set size [m): 3 Number of cydes {r]: 3 Required number of samples (m x r]: 9 Number of field locations to rank [m x m x r): 27 For ranked set sampling . I would need to measure [m x m x r1=27 field locations. raoong them in sets of m:3 using professional judgment . I will coUect r=3 cydes of data . [m x r)=9 samples wm be analyzed i1 the laboratory.

Figure 4.1. Sample Size Determination Using VSP Rather than implementing an RSS sampling design in the Containment Buildings, a simple random sampling approach was adopted. The number of samples was sufficient to meet the objectives outlined in Section 3. The number of measurements collected from the Undervessel Area and 565-ft elevation was 9 and 10, respectively, for each Containment Building.

4.4 ISOCS MEASUREMENTS ISOCS measurements were performed at the randomly selected locations in each of the Containment Building SUs. The locations were planned and mapped using VSP. Judgmental measurement locations were selected based on professional judgement or as identified from surface scans. The detector was positioned and collimated such that the field of view (POV) was approximately 28 m 2 to coincide with the FSS measurement FOV. The FOV was adjusted depending on detector offset limitations due to physical boundaries of the Undervessel Area structure.

Measurements were performed using a portable broad energy high-purity germanium (HPGe) detector. Data acquisition was performed via Canberra's Genie 2000 software. Efficiency curves-Zion Containment and Auxiliary Buildings 14 5271-SR-03-0

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s;: ~ FOR SCIENCE AND EDUCATION encompassing the applicable ROC-for the measurement geometry were generated using Canberra's ISOCS calibration software. The efficiency curves were modeled using ZS's RO C depth profile that was based on applicable characterization data. Specific modeling parameters are discussed in Appendix C.

4.5 VOLUMETRIC SAMPLES Concrete samples were collected from randomly and judgmentally selected locations using a concrete hole-saw and an electric drill. Concrete samples were collected from a depth of up to 15 centimeters in the Undervessel Areas. For the Auxiliary Building, concrete samples were collected from a depth of approximately 6.7 centimeters.

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 E nvironmental Analytical Laboratory (REAL) in Oak Ridge, Tennessee. Sample analyses were performed in accordance with the ORA U Radiological and EnvironmentalAna!Jtical Laboratory Procedures Manual (ORAU 2017).

Concrete samples were analyzed by gamma spectrometry for gamma-emitting fission and activation products. Concrete samples were processed by wet chemistry and material oxidation 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 picocuries per gram pCi per meter squared (pCi/m2) based on the concrete sample depth. Measurement results from the in situ gamma spectrometry measurements were reported in units of pCi/ m 2 . ProUCL, version 5.1 , was used to calculate the UCL95 for both the confirmatory and FSS data set. The mean ROC concentration and associated 95% confidence level were plotted for direct comparison.

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

6.1 UNIT 1 CONTAINMENT 6.1.1 Unit 1 Undervessel Area 6.1.1.1 Surface Scans Overall, NaI detector response in the Undervessel Area ranged from approximately 31,000 to 710,000 counts per minute (cpm). Figure 6.1 provides a quantile plot (Q-plot) of the electronically captured scan data. A review of the Q-plot indicates two populations are present: one in the in-core tunnel and the other in the cylindrical portion of the survey unit-directly under the former reactor pressure vessel (RPV). A sharp increase in the Q-plot is due to two localized areas of elevated activity ["hot spots"] that were identified in the area under the former RPV-one on the wall and the other on the floor. T hese locations were less than 1 m 2 in size and were marked, based on an increase in ratemeter output relative to localized background, for follow-up measurements with the HPGe detector.

I:

i-i I_

z Figure 6.1. Q-plot for Gamma Scan Data of Unit 1 Undervessel Area Floor and Lower Walls Zion Containment and Auxiliary Buildings 16 5271 -SR-03-0

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~ ~ FOR SCLEN CE AND EDUCATLON 6.1.1.2 In Situ Gamma Spectrometry ROC Concentrations Twelve in situ gamma spectrometry measurements- including 9 random and 3 judgmental measurements-were collected from the Undervessel Area. Individual in situ gamma spectrometry measurements collected from this area are presented in Appendix B, Table B-1. Measurement locations from this area are depicted in Figure A-1. Table 6.1 provides a summary of the confirmatory in situ gamma spectrometry measurements.

Table 6.1. Summary of Unit 1 Undervessel Confirmatory In-Situ Gamma Spectrometry Measurements Parameter (pCi/m2) Fractionb ROC Mean Median SD Min Max U CL95* Op. BC Co-60 2.15E+06 8.70E+OS 2.09E+06 1.23E+OS 5.03E+06 5.19E+ 06 0.03 0.03 Cs-134 -6.99E+03 -2.48E+03 2.19E+04 -4.41E+04 2.61E+04 2.48E+04 0.00 0.00 Cs-137 1.56E+06 7.08E+OS 1.88E+06 1.60E+OS 5.20E+06 4.29E+06 0.11 0.11 Eu-152 1.52E+ 07 4.37E+06 1.64E+07 -8.06E+04 3.78E+07 3.90E+07 0.11 0.11 Eu-154 9.58E+ OS 6.64E+OS 9.43E +OS -4.47E+04 2.24E+06 2.33E+06 0.01 0.01 SOF 0.26 0.26

UCL 1s based on the Chebyshev Inequality bQp. represents the UCL95 divided by the Operational DCGL; BC represents the UCL95 divided by the Base Case DCGL SD = standard deviation As indicated in Table 6.1, Cs-137 and Eu-152 contribute the largest fractions towards the mean SU SOP. All individual in situ measurements were less than the DCGL0p. The judgmental measurements collected from this area are relatively close to the UCL95 and less than the DCGL0p; thus, further evaluation is unnecessary as the elevated locations will be represented by the UCL95 in the subsequent analysis .

A comparison of the ORISE and the PSS data mean ROC concentrations and associated uncertainties are provided in Figure 6.2. The error bars in Figure 6.2 represent the uncertainty in the mean concentration, where the upper end is simply the UCL95. For both data sets, the UCL95 was calculated based on the Chebyshev inequality. The data set for Cs-134 and Eu-154 had few detects; thus a direct comparison is unnecessary at concentrations that are a small fraction of the DCGL and near the MDC.

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~ ~ FOR SCIENCE AND EDUCATION Co-60 Cs-134 Cs-137 Eu-152 Eu-154 Be+07 6e+07

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ORJSE Zion ORISE Zion ORISE Zion ORISE Zion ORJSE Zion Figure 6.2. Comparison of PSS Data and ORISE Confirmatory Mean Concentrations and Uncertainties for Gamma-emitting Radionuclides in the Unit 1 Undervessel Area As Figure 6.2 indicates, all mean fractions and their associated confidence intervals overlap for each gamma-emitting ROC. There is a slight positive bias in the ORISE Co-60 mean fraction relative to the FSS mean and a more pronounced positive bias for Eu-1 52. For Cs-137, there is a slight negative bias. These statistical biases are likely attributable to the difference in ISOCS modeling parameters. ORISE does not have the site-specific ISOCS modeling parameters used for FSS measurements. As such, a comparison of the parameters to potentially identify the source of the bias cannot be performed. However, given the overall magnitude of the Operational and Base Case DCGLs relative to ROC concentrations and the associated uncertainty, the additional evaluation of the ISOCS modeling is unwarranted.

The mean SOF, based on the DCGLBc, for the ORISE confirmatory measurements of gamma-emitting radionuclides was 0.26 compared to 0.13 for the FSS data. The factor of two difference between the SU SOF is related to the fact that the ORISE value is based on the U CL95 of the mean SOF, whereas the FSS SOF is based on the arithmetic mean.

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OAK RI DGE INSTITUTE Si I FOR SC IENCE AND EDUCATION 6.1.1.3 ROC Concentrations in Concrete Samples Nine concrete samples-including 8 random samples and 1 judgmental sample-were collected from the Unit 1 Undervessel Area. Individual results for these volumetric concrete samples are provided in Appendix B, Table B-2. The concrete samples were collected from the in situ gamma spectroscopy measurement locations presented in Figure A-1. Concrete samples were not collected from random in-situ measurement location U1 UV-08 and judgmental locations U1UV-10J and -12J, due to temporal limitations. Location U1 UV-11J was selected for concrete sampling because this location exhibited the highest Nal response of the three judgmental in situ measurements.

A summary of the concrete sample concentrations are provided in Table 6.2. All individual ROC concrete sample concentrations were less than the DCGL0 r, with the exception of H-3. Four samples that were collected from in situ measurement locations 5271 U1 UV-02, -03, -04, and -11 j (See Figure A-1) exceeded the DCG~c for H-3.

Table 6.2. Summary of Unit 1 Undervessel Random Volumetric Concrete Samples Parameter (pCi/m2)

ROC Mean Median SD Min Max UCL9Sa Co-60 1.13E+06 2.74E+05 1.88E+06 8.95E+03 5.54E+06 4.03E+06 Cs-134 2.10E+03 9.73E+02 3.39E+03 -3.89E+02 9.99E+03 7.33E+03 Cs-137 1.40E+05 2.94E+04 1.79E+05 3.94E+03 4.62E+05 4.16E+05 Eu-152 1.51E+07 3.52E +06 2.57E+07 1.61E+04 7.56E+07 5.47E+07 Eu-154 5.82E+05 1.13E+05 1.07E+06 -5.37E +03 3.13E+06 2.23E+06 H-3 2.28E+08 5.18E+07 3.75E+08 1.29E+07 1.12E+09 8.09E+08 Ni-63 2.46E+06 3.44E+05 5.09E+06 1. 75E+05 1.49E+07 1.03E+07 Sr-90 -4.48E+02 -7.16E+03 3.58E+04 -3.94E+04 5.73E+04 5.47E+04

UCL95 1s based on Chebyshev's 111equality 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 Undervessel Area along with the corresponding fractional contribution to the SOF from each ROC. The H-3 fraction was 3.40 based on the DCGL8 c (3.41 based on the DCGL0 r) and the UCL95 was approximately 3.5 times higher than the mean concentration, indicating both a high population variability and sampling uncertainty. However, the calculated H-3 fraction was approximately 0.96, based on the arithmetic mean concentration and the DCGL8 c- As presented in Section 3.7.1, it is ORISE's opinion that the Zion Containment and .Auxiliary Buildings 19 5271 -SR-03-0

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~ ~ FOR SClENCE AND EDUCATION UCL95 is the more appropriate parameter for assessing the confirmatory survey data set. As the UCL95 accounts for the uncertainty in the estimate of the mean due to sampling error, the probability of a Type I decision error is minimized (i.e., incorrectly concluding that the H-3 concentration is less than DCGLuJ.

Table 6.3. Unit 1 Undervessel Results by Analysis Method Bv Volumetric Bv ISOCS ROC Concentration* Fractionb Concentration* Fractionb (pCi/m 2) Op. BC (pCi/m 2) Op. BC Co-60 4.03E+06 0.03 0.03 5.19E+06 0.03 0.03 Cs-134 7.33E+03 0.00 0.00 2.48E+04 0.00 0.00 Cs-137 4.16E+05 0.01 0.01 4.29E+06 0.11 0.11 E u-152 5.47E+07 0. 15 0.15 3.90E+07 0.11 0.11 E u-154 2.23E+06 0.01 0.01 2.33E+06 0.01 0.01 Gamma-emitting SOP 0.19 0.19 0.26 0.26

__ d H-3 8.09E+08 3.41 3.40 -- --

Ni-63 1.03E+07 0.00 0.00 -- -- --

Sr-90 5.47E+04 0.04 0.04 -- -- --

SOP 3.65 3.63 -- --

Reported concentration is the UCL95 based on Chebyshev's inequality bQp. represents the UCL95 divided by the Operational DCGL; BC represents the UCL95 divided by the Base Case DCGL cDiscrepancy in summation due to rounding dinferred concentration is not applicable Table 6.3 indicates an agreement between the gamma-emitter mean concentrations determined by both analytical methods for Co-60, Cs-134, and Eu-154. RO C-specific fractions for these radionuclides are agreeable to two decimal places . The Cs-137 fraction is biased high for the ISOCS method compared to the volumetric samples. For Eu-152, the fraction determined by the volumetric samples is slightly higher than that determined by ISOCS. As noted in Section 4.5, the volumetric samples represent a thickness of 15 centimeters, whereas the ISOCS modeling accounts for contamination significantly deeper. Refer to Section C.3.2 for a discussion of the RL determined for each ROC. The volumetric sample depth of six inches accounts for at least 85% of the residual Co-60, Cs-13 7, and Eu-152 activity and 75% of the E u-154 activity modeled with ISO CS. Therefore, based on the magnitude of the relaxation length (RL) for the ROC depth profiles relative to the volumetric sampling depth the volumetric concentrations should be slightly less than those Zion Containment and Auxiliary Buildings 20 5271-SR-03-0

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~ I FOR SCLENCE AND EDUCATION determined by ISOCS . However, this is not the case and there are numerous factors related to each analysis method that account for the difference. Additional investigation into the difference between the Cs-13 7 and E u-152 fractions by the two analysis methods are discussed below.

Analysis by in situ gamma spectrometry overestimates the Cs-137 inventory and corresponding contribution to the SOF by approximately an order of magnitude. Determining the cause of the drastic difference between the two Cs-137 inventories would require the collection of additional data. It is reasonable to assume that, due to the extensive remediation, the residual Cs-137 activity is not uniform over the HPGe detector FOV-both in terms of width and depth. The HPGe FOV represents an area much larger than the volumetric samples and will smooth out the spatial variability. If the volumetric sample does not represent the spatial variability within the HPGe FOV, then the results will be significantly different. Nevertheless, the volumetric samples provide evidence that the ISOCS measurement modality is conservative for Cs-137.

T he Eu-152 UCL95 mean concentration determined by volumetric sampling is approximately 40%

higher than the concentration determined by the in situ measurements. Additional data would be required to determine the cause of the difference. To simplify the comparison of the two analytical method results and account for the corresponding under- and over-estimate of doses to a potential future receptor, the combined mean Cs-137 and Eu-152 volumetric sample concentration fraction of 0.08 was less than the mean of 0.11 for the in situ measurements.

6.1.2 Unit 1 Containment 565-ft Elevation 6.1.2.1 Surface Scans Overall direct gamma radiation ranged from 3,300 to 74,000 cpm, inclusive of the two sumps.

Figure 6.3 depicts the quantile-quantile (Q-Q) plot for the 565-ft E levation and Above. As depicted in Figure 6.3, scan ranges in the sumps were lower relative to the floors and lower walls. For the two sumps, Nal detector response ranged from 3,300 to 47,000 cpm. The surveyors noted a few locations of elevated direct gamma radiation on the top of the sump where the wall met the 565-ft floor elevation. These locations were discrete Oess than 100 cm2) and did not warrant further investigation.

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~ ~ FOR SCIENCE AND EDUCATION Unit 1 565 ft Elevation e """

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Figure 6.3. Q-Q Plot for Gamma Scan D ata of Unit 1 565-ft Elevation Nal detector response for the balance of the 565-ft elevation, consisting of the floor and lower walls, ranged from 4,400 to 74,000 cpm. Several discrete areas of elevated direct gamma radiation were identified on the ridges of the lower walls. There appeared to be residual concrete from demolition activities associated with these anomalies. Figure 6.4 depicts an example of the elevated areas previously discussed. Given the relatively small area of elevated direct gamma radiation relative to the total surface area of the SU, additional investigation was determined unnecessary as the contribution to the ROC concentrations from these elevated locations will be represented by the in situ gamma spectrometry measurements.

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~ I FOR SCIENCE AND EDUCATION Figure 6.4. Example Location of Elevated Direct Gamma Radiation Identified on Lower Walls of the Unit 1565-ft Elevation 6.1.2.2 In Situ Gamma Spectrometry Measurements A total of 10 random in situ gamma spectrometry measurements were collected from the Unit 1 565-ft Elevation. A summary of the individual gamma spectrometry measurements collected in this area is provided in Appendix B, Table B-3. The measurement locations are depicted in Figure A-2.

Applicable summary statistics for this SU are presented in Table 6.4. Cs-137 and Co-60 were the only radionuclides identified above their respective MDC. The maximum measurement for each detected ROC was compared directly to the DCGL0p, resulting in a fraction of 0.06 and 0.00 for Cs-137 and Co-60, respectively. The mean SOF for this SU is 0.01 based on the UCL95 and DCGLBc, which is equivalent to the SOF for gamma-emitting radionuclides presented in the FSS data.

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~ I FOR SCIENCE AND EDUCATION Table 6.4. Summary of Unit 1565 ft Elevation Confirmatory In-Situ Gamma Spectrometry Measurements Parameter (pCi/m2) Fractionh ROC Mean Median SD Min Max UCL95* Op. BC Co-60 3.43E+04 3.72E+ 04 1.03E+04 1.96E+04 5.14E+04 4.85E+04 0.00 0.00 Cs-134 1.56E+03 5.66E+02 8.59E+03 -1.18E+04 1.90E+04 1.34E+04 0.00 0.00 Cs-137 1.30E+05 9.10E+04 1.00E+OS 1.29E+04 3.16E+05 2.68E+05 0.05 0.01 Eu-152 9.65E+03 7.35E+03 2.50E+04 -3.96E +04 5.66E+04 4.42E+04 0.00 0.00 Eu-154 -2. 19E+04 -2.35E+04 4.34E+04 -1 .07E+05 4.57E+04 3.79E+04 0.00 0.00 SOFc 0.06 0.01

UCL95 is based on the Chebyshev Inequality hQp. represents concentration fraction of the Operational DCGL; BC represents concentration fraction of the Base Case DCGL cDiscrepancy in summation due to rounding SD = standard deviation Gamma-emitting ROC mean concentrations and their associated 95% confidence intervals for both ORISE and the PSS data are plotted in Figure 6.5. As indicated in Figure 6.5, all ROC mean concentrations overlap at the 95% confidence level although for each ROC, the PSS concentration estimates are biased high relative to those determined by ORISE. This observed bias may be result of several measurements in the PSS sample population that were collected near the Undervessel Area cavity opening, such that radionuclides in that concrete contributed additional counts to the acquired PSS in situ spectrum.

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~ ~ FOR SClENCE AND EDUCATION Co-60 Cs-137 Cs-143 Eu-152 Eu-154 3e+OS 1u I

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-1e+OS ORI SE FSS Data ORISE FSS Data ORISE FSS Data ORISE FSS Data ORISE FSS Data Figure 6.5. Comparison of PSS Data and ORI SE Confirmatory Mean Concentrations and Uncertainties for Gamma-emitting Radionuclides in the Unit 1565-ft Elevation 6.2 UNIT 2 CONTAINMENT 6.2.1 Unit 2 Undervessel Area 6.2.1.1 Sur.ii. cc Scans Nal detector response in the Undervessel Area ranged from approximately 21,000 to 480,000 cpm.

Figure 6.6 provides a Q-plot of the electronically captured data. A review of the Q-plot indicates two populations are present: one in the in-core tunnel and the other in the cylindrical portion of the area-directly under the former RPV. No judgmental locations were selected for follow up measurement with the HPGe detector based on the respective uniform detector responses in the in-core tunnel and cylindrical area.

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~ ~ FOR SC I ENCE AND EDUCATIO N

.. ,* * * *** I

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Figure 6.6. Q-plot for Gamma Scan D ata of U nit 2 Undervessel Area Floor and Lower Walls 6.2.1.2 In Situ Gamma Spectrometry Measurements A total of nine random in situ gamma spectrometry measurements were collected from the U nit 2 Undervessel Area. Individual gamma spectrometry measurements collected from this area are presented in Appendix B, Table B-4. Measurement locations from this area are depicted in Figure A-3. Table 6.5 provides a summary of the confirmatory in situ gamma spectrometry measurements. All individual in situ measurements were below the Operational and Base Case D CGLs.

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~ I FOR SCl ENCE AND EDUCATIO N Table 6.5. Summary of Unit 2 Undervessel Confirmatory In-Situ Gamma Spectrometry Measurements Parameter (pCi/m2) Fractionh ROC Mean Median SD Min Max UCL95* Op. BC Co-60 3.07E+06 1.07E+ 06 3.41E+06 7. 62E+03 8.99E+06 8.02E+06 0.05 0.05 Cs-134 -9.63E+03 -1.42E+03 4.00E+04 -8.80E+04 6.34E+04 4.85E+04 0.00 0.00 Cs-137 1.37E +06 6.21E+05 1.64E+06 2.25E+04 5.04E+06 3.75E+06 0.10 0.10 Eu-152 1.65E+07 4.89E+ 06 1.81E+07 -2.62E+04 4.41E+07 4.27E+ 07 0.12 0.12 Eu-154 1.04E+06 9.81E+05 1.03E+06 -7.01E+04 2.31E+06 2.54E + 06 0.01 O.Ql SQFc 0.27 0.27

UCL95 is based on the Chebyshev Inequality bQp. represents the UCL95 divided by the Operational D CGL; BC represents the UCL95 divided by the Base Case DCGL cDiscrepancy in summation due to rounding SD = standard deviation As indicated in Table 6.6, Cs-137 and E u-152 are the largest gamma-emitting ROCs to the SU SOF.

A comparison of the ORJSE and the FSS data mean ROC concentrations and associated uncertainties is provided in Figure 6.7. The error bars in Figure 6.7 represent the uncertainty in the mean concentration, where the upper end is simply the UCL95. For both data sets, the UCL95 was calculated based on the Chebyshev inequality. The data set for Cs-134 and Eu-154 had few detects, thus a direct comparison is unnecessary for these concentrations that were a small fraction of the DCGL and near the MDC.

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~ ~ FOR SCIENCE AND EDUCATION Co-60 Cs-134 Cs-137 Eu-152 Eu-154 8e+07

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ORJSE Zion ORISE Zi on ORJSE Zion ORJSE Zion ORISE Zion Figure 6.7. Comparison of PSS Data and ORISE Confirmatory Mean Concentrations and Uncertainties for Gamma-emitting Radionuclides in the Unit 2 Undervessel Area As Figure 6.7 indicates, all gamma-emitting ROC fractions and their associated confidence intervals overlap for the confirmatory and FSS populations. There is a slight positive bias in the ORISE Co-60 fraction relative to the FSS result and a more pronounced positive bias for Eu-152. For the OR1SE Cs-137 result, there is a slight negative bias.

The gamma-emitter SOF calculated from confirmatory measurements, based on the DCG~c, was 0.27 compared to 0.09 reported for the FSS data; the higher confirmatory SOF also resulted from using the UCL95 of the mean SOF versus an arithmetic mean as discussed for Unit 1 Containment.

As with the in situ gamma spectrometry measurements in the Unit 1 Undervessel Area, the slight biases in the arithmetic mean SOF are likely attributable to the difference in ISOCS modeling parameters.

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~ ~ FOR SCIENCE AND EDUCATION 6.2.1.3 ROC Concentration in Concrete Samples A total of 8 random concrete samples were collected from the Unit 2 Undervessel Area. Individual results for volumetric concrete samples collected from this area are provided in Appendix B, Table B-5. The concrete samples were collected from the in situ measurement locations shown in Figure A-

3. A summary of the concrete sample concentrations is provided in Table 6.6. Three of the nine planned concrete samples were not collected as all concrete had been removed at these locations, leaving only the steel liner. All individual concrete sample concentrations were less than the DCGL0p, with the exception of one sample for H-3. This sample was collected from in situ measuremen t location 527 1U2UV-03 (See Figure A-3).

Table 6.6. Summary of Unit 2 Random Volumetric Concrete Samples Parameter (pCi/ m2)

ROC Mean Median SD Min Max UCL95*

Co-60 7.05E+05 5.86E+05 6.69E+05 1.01E+05 1.78E+06 1.90E+06 Cs-134 1.95E+03 1.62E+03 2.31E+03 -4.09E+02 5.84E+03 6.06E+03 Cs-137 1.05E+05 3.46E+04 1.80E+05 7.16E+03 4.69E+05 4.25E+05 Eu-1 52 8.92E+06 7.65E+06 8.27E+06 1.35E+06 2.05E+07 2.36E+07 E u-154 3.75E+05 2.80E+05 4.05E+05 1.79E+04 1.05E+06 1.10E+06 H-3 1.29E+08 1.06E+08 9.38E+07 3.78E+07 2.69E+08 2.96E+08 Ni-63 2.58E+05 2.81E+05 6.01E+04 1.68E+05 3.15E+05 3.65E+05 Sr-90 5.97E+03 -1.79E+03 3.16E+04 -2.51E+04 5.01 E+04 6.22E+04

UCL95 is based on Chebyshev's inequality SD = Standard deviation Table 6.7 provides the ROC concentration determined by both the volumetric samples and by the in situ gamma spectrometry measurements for the Undervessel Area along with the corresponding SOF. All ROC concentrations were a fraction of the DCGL8 c, by both analysis methods, with the exception of H-3, which has a fraction relative to the DCGL8 c of 1.25. The UCL95 is approximately 2.3 times higher than the mean concentration, indicating both a high population variability and sampling uncertainty. The H-3 fraction is approximately 0.54 if the H -3 concentration arithmetic mean is used in the calculation. As previously discussed, it is ORISE's opinion that the UCL95 is the more appropriate parameter for assessing the confirmatory survey data set.

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~ ~ FOR SCIENCE AND EDUCATION Table 6.7. Unit 2 Undervesssel Results by Analysis Method By Volumetric By !SOCS ROC Concentration* Fractionb Concentration* Fractionb (pCi/m ) 2 (pCi/m ) 2 Op. BC Op. BC Co-60 1.90E+06 0.01 0.01 8.02E+06 0.05 0.05 Cs-134 6.06E+03 0.00 0.00 4.85E+04 0.00 0.00 Cs-137 4.25E+05 0.01 0.01 3.75E+06 0.10 0.10 Eu-152 2.36E+07 0.06 0.06 4.27E+07 0.12 0.12 Eu-154 1.10E+06 0.00 0.00 2.54E +06 0.01 0.01 Gamma-emitting SO.Fe 0.09 0.09 0.27 0.27 H-3 2.96E+08 1.25 1.25 -- d -- --

Ni-63 3.65E+05 0.00 0.00 -- -- --

Sr-90 6.22E+04 0.04 0.04 -- -- --

SO.Fe 1.39 1.39 -- --

Reported concentration is the UCL95 based on Chebyshev's inequality hOp. represents the UCL95 divided by the Operational DCGL; BC represents the UCL95 divided by the Base Case DCGL cDiscrepancy in summation due to rounding dinferred concentration is not applicable Table 6.7 indicates that the ROC concentrations determined by the in situ gamma spectrometry measurements is more conservative [higher] than that determined based on the volumetric samples.

As with Unit 1, determining the cause of the difference would require additional data, however, such evaluation in unnecessary due to the conservative nature of the in situ measurements.

6.2.2 Unit 2 Containment 565-ft Elevation 6.2.2.1 Surface Scans Overall direct gamma radiation ranged from 3,000 to 46,000 cpm, inclusive of the two sumps.

Figure 6.8 depicts the Q-Q plot for the floor, lower walls, and sumps. As depicted in Figure 6.8, scan ranges in the sumps were less than levels associated with the floor and lower walls. For the two sumps, Nal detector response ranged from 3,000 to 6,000 cpm.

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/

.I Figure 6.8. Q-Q Plot for Gamma Scan Data of Unit 2 565-ft Elevation NaI detector response for the balance of the 565-ft Elevation, consisting of the floor and lower walls, ranged from 4,100 to 46,000 cpm. Several discrete areas of elevated direct gamma radiation were identified on the ridges of the lower walls. As discussed in Section 6.1.2 and shown in Figure 6.4 for Unit 1, residual concrete from demolition activities was also present on the ridges of the Unit 2 walls and as with Unit 1, additional investigation was determined unnecessary as the contribution to the RO C concentrations from these elevated locations will be represented by the in situ gamma spectrometry measurements.

6.2.22 In Situ Gamma Spectrometry Measurements A total of 10 random in situ gamma spectrometry measurements were collected from the Unit 2 565-ft Elevation. A summary of the individual gamma spectrometry measurements collected from this area are provided in Appendix B, Table B-6. Measurement locations are shown in Figure A-4.

Applicable summary statistics for this SU are presented in Table 6.8. Cs-137 and Co-60 were the only radionuclides identified above their respective l\IDC. The maximum measurement for each detected ROC was compared directly to the DCGL0 r, resulting in a fraction of 0.01 and 0.00 for Cs-Zion Containment and Auxiliary Buildings 31 5271-SR-03-0

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~ ~ FOR SCIENCE AND EDUCATION 137 and Co-60, respectively. The mean SOP for this SU is 0.00, based on the UCL95 and DCG:Lnc, which is equivalent to the PSS reported SOF for gamma-emitting radionuclides.

Table 6.8. Summary of Unit 2 565-ft Elevation Confirmatory In-Situ Gamma Spectrometry Measurements Parameter (pCi/m2) Fractionb ROC Mean Median SD Min Max UCL95* Op. BC Co-60 2.12E+ 04 2.02E+04 8.85E+03 1.03E+04 3.80E+04 3.34E+04 0.00 0.00 Cs-134 3.49E+03 4.01E+03 5.68E+03 -4.77E+03 1.08E+04 1.13E+04 0.00 0.00 Cs-1 37 3.4SE+04 2.92E+04 1.71E+04 2.37E+04 7.93E+04 5.81E+04 0.01 0.00 Eu-152 -2.43E+03 -8.12E+03 1.33E+04 -1.66E+04 1.94E+04 1.60E+04 0.00 0.00 Eu-154 -8.63E+03 -9.32E+03 2.20E+04 -4.13E+04 2.79E+04 2.17E+04 0.00 0.00 SOFc 0.02 0.00

UCL95 is based on the Chebyshev Inequality hOp. represents concentration fraction of the Operational DCGL; BC represents concentration fraction of the Base Case DCGL cDiscrepancy in summation due to rounding SD = standard deviation

Gamma-emitting ROC mean concentrations and their associated 95% upper confidence level for both the ORlSE and FSS data are plotted in Figure 6.9. As indicated in Figure 6.9, all RO C mean concentrations overlap at the 95% confidence level. All mean ROC FSS concentrations are biased high relative to confirmatory concentration results, similar to the bias observed in the data comparison for the Unit 1 565-ft Elevation, and also likely the result of Undervessel Area cavity opening gamma interference to some of the acquired FSS measurements .

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~ ~ FOR SClENCE AND EDUCATION Co-60 Cs-137 Cs-143 Eu-152 Eu-154 2e+05 "E

"'0

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DRISE FSS Data Figure 6.9. Comparison of FSS Data and ORI SE Confirmatory Mean Concentrations and Uncertainties for Gamma-emitting Radionuclides in the Unit 2 565-ft Elevation 6.3 AUXILIARY BUILDING 6.3.1 Surface Scans Overall Nal detector response ranged from approximately 7,000 cpm to over 1,000,000 cpm, inclusive of the sumps. Figure 6.10 presents the Q-plot for surface scans of the floor and sumps.

The flat line at 1,000,000 cpm represents the maximum range of the ratemeter-scaler operated in rate mode. The survey team noted several embedded piping/ penetrations in the floor and sump that were previously grouted and exhibited relatively high direct gamma radiation. These locations were assessed as part of the site's FSS of embedded piping/ penetrations prior to grouting and, therefore, excluded from further confirmatory assessment. Excluding the aforementioned embedded piping/penetrations, four locations with direct gamma radiation levels higher than 900,000 cpm, representing the maximum observed gamma radiation locations, were identified and selected for additional investigation as a result of the surface scans.

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Si ~ FOR SCLENCE AND EDUCATION AIU 8u,khna Roo, end Sur.pa

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Figure 6.10. Q-plot for the Auxiliary Building Floor and Sumps At the request of the NRC, ORISE surveyed the lower portion (below 6 ft) of the of the west wall in the Auxiliary Building. NRC requested that ORISE focus on the northern most portion of the west wall, in the former HUT cubicles, therefore, the upper portion (above 6 ft) of the west wall in this area was scanned. Direct gamma radiation ranged from approximately 8,400 to 900,000 cpm. NRC requested that ORISE mark and then collect a sample from the location in the former northern most HUT cubicle exhibiting the highest gamma radiation level, which was 900,000 cpm, for comparison to the site's volumetric sample results from the same area. The location exhibiting the highest direct gamma radiation was localized without the ratemeter-scaler electronically capturing data and, thus, not included in Figure 6.11 .

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si ~ FOR SCIENCE AND EDUCATION AwK auikfine w..tw.el Figure 6.11. Q-plot for the Auxiliary Building West Wall 6.3.2 ROC Concentrations in Volumetric Samples Fifteen concrete samples-including nine random and six judgmental samples-were collected from the Auxiliary Building. Individual results for these samples are provided in Appendix B, Table B-7.

The concrete sampling locations are provided in Figure A-5. A summary of the randomly selected concrete sample concentrations is provided in Table 6.9. All individual sample results were below the Operational and Base Case DCGLs, with the exception of judgmental samples M0016 and M0017, which were collected from the edge of floor and the Northern most HUT cubicle wall, respectively. Both judgmental samples [M0016 and M0017] exhibited Cs-137 concentrations above the DCGL8 c*

Table 6.9. Summary of ROC Concentrations in the Auxiliary Building Floor Parameter (pCi/ m 2)

ROC Mean Median SD Min Max UCL95*

Cs-137 2.25E+05 8.04E+04 3.04E+05 1.22E+04 9.36E+05 6.52E+05 Cs-134 2.52E+03 2.19E+03 2.93E+03 -1.88E+03 7.05E+03 6. 78E+03 Co-60 1.27E+05 8.31E+03 3.63E+05 -2.04E+03 1.09E+06 6.67E+05

__ b Sr-90 -- -- -3.60E+ 04 5.74E+05 --

Ni-63 -- -- -- 3.02E+05 1.1 7E+09 --

Based on the Chebyshev inequality bSummary statistics are not applicable as the entire random sample set was not analyzed Zion Containment and Auxiliary Buildings 35 5271 -SR-03-0

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~ ~ FOR SCLENCE AND EDUCATION The two judgmental samples that had a Cs-137 concentration in excess of the DCGLsc [M0016 and M0017] were collected from locations that had a fairly localized direct gamma radiation (less than 1 m2). The FSS measurements-by ISOCS-for both of these locations were 1.98E+07 pCi/m2 and 3.75E+06 pCi/m2 for the HUT Cubicle west wall and floor, respectively. The FSS measurements represent the ROC activity concentration averaged over an area of 28 m 2, whereas the volumetric concrete sample represents an area of 0.005 m 2.

Judgmental samples M0020 and M0021 were collected from the same location, which was near Sump Pit B. This location had sand/ sediment on top of the concrete. The sand/ sediment was collected as sample M0020 while the concrete was collected as sample M0021. When collecting the sand/ sediment portion [M0020], the surveyors noted that the residual radioactivity contributing to the 900,000-plus gamma cpm was obtained in the sample as indicated by a sharp decrease in the direct gamma radiation levels at the location post-sample collection. \Vhile processing sample M0020, REAL staff were able to separate a hot particle from the sample. The particle is depicted in Figure 6.12. Gamma spectroscopy identified approximately 8 µCi of Co-60 as the sole ROC in the hot particle., The reported activity is a semi-quantitative value due to gamma spectrometer dead time

[greater than 90%) and because an appropriate calibration geometry was not available.

Figure 6.12. Hot Particle Separated from Concrete Sample M0020 Concrete sample M0018 was collected from a location that exhibited a direct gamma radiation level over one million cpm. However, the ex situ direct gamma radiation level on the collected sample was Zion Containment and Auxiliary Buildings 36 5271 -SR-03-0

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~ ~ FOR SCIENCE AND EDUCATlON unremarkable. Additionally, the direct gamma radiation level of the sample area sharply decreased post sample collection. ROC concentrations in sample M0018 were relatively low compared to other samples that exhibited high detector response (samples M0016, MOOl 7, and M0020 exhibited high direct gamma radiation and a commensurately high ROC concentration). A possibility is that there was another hot particle present at this location but was not collected. A significant amount of water was present that surrounded this sampling location, thereby providing a migration path from the sampling area.

The sample collected from the west wall [MOOl 7] exhibited elevated ROC concentrations; primarily Cs-137 at 9.32E+08 pCi/ m 2

ZS had previously collected abiased sample from this area. ZS's sample result was approximately 5.51E+08 pCi/ m 2 (averaged over a three-inch sample depth),

which is approximately 40% lower than the ORISE sample result.

Six samples were selected for hard-to-detect (HTD) analysis to evaluate the surrogate ratios specified in the LTP . ROC HTD concentrations and their corresponding surrogate ratios are presented in Table 6.10. The maximum Ni-63:Co-60 and Sr-90:Cs-137 surrogate ratios were 52.96 and 0.0004, respectively, which are significantly less than the values specified in the LTP.

Table 6.10. Summary ofHTD ROC Concentrations in Select Auxiliary Building Concrete Samples Ni-63 Co-60 Sr-90 Cs-137 Sample ID Ni:Co Sr:Cs (pCi/~) (pCi/~) (pCi/!!) (pCi/~)

__ a 5271M0009 30.3 6.98 4.34 0.06 3.09 5271M0015 1.93 0.048 40.2 -0.23 5.97 --

5271M0016 7478 141.2 52.96 3.66 8870 0.0004 5271M0017 285 9.16 31.1 2.66 5950 0.0004 5271M0020 407 34.3 11.87 0.26 140 --

5271M0021 151 .2 88 1.7 0.2 56.3 --

Surrol[ate Ratios from LTP, Chapter 5 Table 5-15 Ni-63:Co-60 180.45 Sr-90:Cs-13 7 0.002

Indicates that the surrogate rauo could not be calculated because Sr-90 was not 1dent1fied above the analytical MDC At the request of the NRC, water sample W0002 was collected from the Auxiliary Building sump prior to draining; results are included in Table 6.11. Once the sump was drained, sediment/water sample M0038 was collected from material remaining inside of the sump. During collection of the sample, the surveyor noted that the sediment material was fine and resuspendable in water. Prior to Zion Containment and .Auxiliary Buildings 37 5271 -SR-03-0

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~ ~ FOR SClENCE AND EDUCATION analysis of sample M0038, the water was filtered and the water and sediment portions were analyzed separately and the results are presented in Table 6.11 .

Table 6.11. ROC Concentrations in Water/Sediment Auxiliary Building Sump Samples Sample ID Cs-137 Cs-134 Co-60 5271W0002a 313 +/- 24c 0.3 +/- 2.2 1.4 +/- 2.3 5271M0038 - Water* 707 +/- 62 -1.4 +/- 3.8 8.1 +/- 4.0 5271M0038 - Sedimentb 647 +/- 55 -0.07 +/- 0.42 284 +/- 17

'Results are 1n uruts of pC1/ liter.

bResults are in units of pCi/ g.

cuncertainties represent the total propagated uncertainty reported at the 95% confidence level.

6.3.3 Auxiliary Building Floor ROC Concentration Assessment The Auxiliary Building floor ROC concentrations were assessed based on the randomly collected concrete samples and represent an unbiased estimate of the mean concentration. For comparison purposes, the FSS ISOCS results for the floor were compared against these confirmatory volumetric sample results for gamma-emitting radionuclides. Figure 6.13 presents a comparison of the mean ROC-concentration and associated 95% confidence level based on the FSS in situ measurements and the confirmatory volumetric sample data.

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~ ~ FOR SCIENCE AND EDUCATION Co-60 Cs-134 Cs-137 1.0e-+-07 N~ 7.5e-+-Q6 E

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O.Oe-+-00 I - -- I FSS ORISE FSS ORJSE FSS ORISE Figure 6.13. Comparison of FSS Data and ORI SE Confirmatory Mean Concentration s and Uncertainties for Gamma-emitting Radionuclides in the Auxiliary Building Floor A review of Figure 6.13 reveals a low bias in the ORISE assessment of the Auxiliary Building floor ROC concentrations. The confidence intervals overlap for Co-60 but not Cs-137; comparison of Cs-134 is urinecessary as the data sets contain numerous non-detects. Numerous factors impact the difference between the FSS Cs-137 concentration and the ORISE-determined concentration and additional data are needed to evaluate this difference. Since the FSS data are biased high relative to the volumetric sample results additional investigation is unwarranted.

As previously discussed, the random samples represent an unbiased estimate of the ROC concrete concentration. Additionally, to assess the potential impact of hot spots on the mean SOF of the Auxiliary Building Floor, the judgmental sample results were also considered. The maximum field ranking measurement for the random RSS sampling locations was approximately 150,000 cpm.

Therefore, it is reasonably conservative to assume that al detector responses below this value are represented by corresponding volumetric samples. The proportion of scan data points above this 150,000 cpm threshold was determined to be 0.10. More directly, 10% of the Auxiliary Building Zion Containment and .Auxiliary Buildings 39 5271 -SR-03-0

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floor area was considered to be composed of "hot spots." This estimate of hot spot area is conservative as grouted penetrations exhibiting elevated gamma radiation levels are also represented in the floor scan data, biasing the scan data high. For the hot spot assessment, the SOF was calculated using Equation 3-1 and assuming that 10% of the Auxiliary Building f}.oor [i.e.

(SA5u/SAEiev) = (1/0.10) in Equation 3-1] had a concentration equal to the geometric mean of the

\

judgmental samples. The geometric mean of the judgmental concrete samples was 3.57E07 pCi/ m 2, 8.24E03 pCi/m2, and 1.34E06 pCi/m2 for Cs-137, Cs-134, and Co-60-respectively. The geometric mean is more appropriate than the arithmetic mean for these results because the geometric mean can provide a better indication of central tendency than the arithmetic mean'. when the data set is skewed, as in this case. Table 6.12 provides a summary of the SOF assessment for the Auxiliary Building floor.

Table 6.12. Assessment of ROC Concentrations in the Auxiliary Building Floor

. ORISE Confirmat_ory Data ,,

ROC Meana Hot Spots Fraction

(pCi/m2) (pCi/m2) Op.b, BCc Cs-137 2.25E+OS 3.57E+07 0.01 0.03 Cs-134 2.53E+03 8.24E+03 0.00 0.00 Co~60 1.27E+OS 1.34E+06 0.00 0.00 aArithmetic mean bCakulated by the arithmetic mean divided by the DCGL0p cCalculated using Equation 3-1

7.

SUMMARY

AND CONCLUSIONS At the NRC's request, ORISE conducted confirmatory survey activities at ZNPS during the period of April 16_,26, 2018. The survey activities included gamma surface scans, in situ gamma spectrometry measurements, and volumetric sampling. Summary and conclusions for each of the areas investigated are summarized below~

7 .1 CONTAINMENT BUILDINGS Based on the overlap of confidence intervals for the in situ gamma spectrometry measurements, ORISE did not identify issues that contradicted the gamma-emitting ROC FSS data for I i demonstrating compliance with the release criterion. All confirmatory in situ gamma spectrometry I

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OAK RIDGE INSTITUTE I I FOR SCIENCE AND EDUCATION measurements in both Containment Buildings were less than the DCGL0 P. ORISE noted that the gamma-emitting SOF determined for the Undervessel Area in each Containment Building-based on the confirmatory measurements-is a factor of two higher than that assigned by ZS. The most direct cause for this difference is due to the fact that ORISE based the SOF on the UCL95 of the

mean but the FSS SOF was based on the arithmetic mean of the data. ORISE elected to use the UCL95 of the mean to account for population uncertainty as the confirmatory survey measurements did not cover the entire SU area, whereas the FSS measurements did. However, employing the UCL95 of the mean over the arithmetic mean does not alter the decision regarding the gamma-emitting ROC concentrations for both Unit 1 and 2 Containment Buildings being less than the NRC-approved limit. The ROC concentrations for the 565-ft elevation in each Containment Building are close to zero, in terms of the SOF, and are unremarkable.

The HTD mean concentration and individual measurements for Sr-90 and Ni-63 were unremarkable. ORISE determined that three individual tritium results from Unit 1 Containment Building and one individual tritium result from Unit 2 Containment Building, exceeded the DCGL8 c. The tritium sample results were based on volumetric samples that represented a depth of six inches. ORISE recommends that NRC evaluate the potential for tritium contamination greater than six inches in the Undervessel Area concrete.

7 .2 AUXILIARY BUILDING Based on the volumetric confirmatory survey sample results, the FSS in situ gamma spectrometry measurements provide a conservative representation of the residual radioactivity inventory. ORISE collected two concrete samples-one from the west wall (northern most HUT Cubicle) and the second from the floor-that had Cs-137 concentrations exceeding the DCGLnc* These tw.o locations were from areas with localized direct gamma radiation. The FSS measurements by ISOCS were less than the DCGL0 P; the difference is due to the in situ gamma spectrometry measurement area being much larger than the area represented by the elevated area from which the concrete sample was collected. The resulting upper bound of confirmatory gamma-emitting ROC SOF for the Auxiliary Building Floor, including hot spots, was a small fractio1:1 relative to the DCGLs. The

HTD results for the Auxiliary building did not indicate that the approved ratios presented in the license termination plan were non-conservative.

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~ ~ FOR SCIENCE AND EDUCATION Confirmatory survey results indicate that there is mobile radioactivity present in the Auxiliary Building Basement-as indicated by the water and sediment sample results. At this time, ORISE cannot confirm the source of the contaminated particulates that were present in the sump.

1*

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8. REFERENCES Canberra 2009. Model S573 !SOCS Calibration S oflware Tech1Jical Reference Manual. Canberra Industries, Inc. Meriden, Connecticut.

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

EPA 2002. Guidance on Choosing a Sampling Design far Environmental Data Collection. EPA QA/ G-5S.

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

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

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

NRC 2000. Multi-Agemy Radiation Survry and Site Investigation Manual (NIARSSIM). NUREG-1575; Revision 1. U.S. Nuclear Regulatory Commission. Washington, D.C. August.

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

ORAU 2016a. ORAU Radiological and Environmental Survry 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. *

.i; ORAU 2016c. ORAU Health and Sefe!J Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. January.

ORAU 2017. ORAtJ Radiological andEnvironmentalAnafytical Laboratory Procedttres Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. August 24.

ORAU 2015. Requestfar Additional ieformation from the Independent Technical Review of Technical Support Document Use ofIn-Situ Gamma Spectroscopy far Source Term Survry ofEnd State Stn1ctures, Zion Nuclear Power Station, Zion, Illinois. 5271-DR-01-0. Oak Ridge Institute for Science and Education. Oak Ridge, Tennessee. October 14.

ZS 2017. Zion Station Restoration Prqject License Termination Plan, Rev. 1. ZionSolutions, LLC. Chicago, Illinois. July.

Zion Containment and Auxiliary Buildings 43 5271-SR-03-0

APPENDIX A: FIGURES Zion Containment and Auxiliary Buildings 5272-SR-03-0

lnooro Floor lrocore TUMII Sol.th waU

l5271U1UV-03!

'"Q D lncore TunMI ceiling Sloped Tunnel cei ling

- r=2==1,!l,1"'"'u1.,.,u=v=

-0.,.,1 Sk>ped Tunruitl floor ZionNPS X

ISOCS Measurement Locations Judgmental ISOCS Locations II ORISE Created by: A. Xi.rtb.li.nk Zion, Illinois ISOCS Measurement Locations for Unit 1 Containment Undervessel n .ue: Jwu 19, 2018

\ ' \ IE.A: Ptott(':tl \ !271 .tiot1 :SPI..

Figure A-1. Unit 1 Undervessel Area In Situ Gamma Spectrometry Measurement Locations Z ion Containment and .Auxiliary Buildings A -1 5272-SR-03-0

l5271U1 CM T-05!

, ****~.

!5271 U1 CMT-02!

\

J5271 U1CMT-06l ;

/<.

/

i~52_7_1_U_1-CM- T--04~!

l5271U1 CMT-08!

I ZionNPS ti EB ISOCS Measurement Locations Zion, Illinois ISOCS Measurement Locations for Unit 1 Containment 565 ft Elev.

and Above Cttwcd by: A. Ki.rdilink Date: June 19, 2018 Figure A-2. Unit 1565-ft Elevation In Situ Gamma Spectrometry Measurement Locations Note: the grayscale does not indicate data related to the confirmatory survey Zion Containment and .Auxiliary Buildings .A-2 5272-SR-03-0

lnoore r.....1East wall lncore TUMtl 't'Utst 11,1aU Undlf'VHHIW~

0lncort Tunnel ca.iling D

Sloped Tcmel ooilng 5271 U2UV-01 Slopod lunnol floor ZionNPS ISOCS Measurement Locations II ORISE Creute:d by:A. Kirthlink Zion, Illinois ISOCS Measurement Locations for Unit 2 Containment Undervessel Date: June: 19 ,2018 Figu re A-3. Unit 2 Undervessel Area In Situ Gamma Spectrometry Measurement Locations Zion Con tainment and Auxiliary Buildings A-3 5272-SR-03-0

i6271U2CMT*051

~

i5271U2C IAT-07!

i5271U2CMT- 10l

!5271 U2C MT-03l

/

.@271U2CMT.()Ql *.._

ZionNPS

!SOCS Measurement Locations II ORISE Zion, Illinois ISOCS Measurement Locations for Unit 2 Containment 565 ft Elev.

Ctt-atcd by: A. Kirthlink.

and Above D :uc:Jw1e 19, 2018 Figure A-4. Unit 2 565-ft Elevation In Situ Gamma Spectrometry Measurement Locations Note: the grayscale does not indicate data related to the confirmatory survey Zion Containment and Auxiliary Buildings A.-4 5272-SR-03-0

Elevator Pit Sump Pit A

- =="'""==

!~271 MOOOS! 1 5271 MD01 9!

l------'lr 5"'"27 _ 1_ M _0'-0'-1_,

6

~271 M0011

@211M001al

..---~

!5271M0014!

.,..,@.,..21=-1..,.M00...,..,,....1"="zj

- ~----.

5271 M0020 & 21

- ~--~

!5271 M0013f

!5271 M0015!

Sump PitB X

Sample Locations Judgmental Location II ORISE Cre ated by: A. Ki rthlink Zion NPS Zion, Illinois Auxiliary Building Sample Locations D rue : Au,:mu 2 1, 2018 Figure A-5. Auxiliary Building Sampling Locations Zion Containment and .Auxiliary Buildings .A-5 5272-SR-03-0

APPENDIX B: DATA TABLES Zion Containment and Auxiliary Buildings 5272-SR-03-0

Table B-1. In Situ Gamma Spectrometry Measurements from the Unit 1 Undervessel Area (pCi/m 2)

Co-60 Cs-134 . Cs-137 Eu-152 Eu-154

Measurement ID Result MDC Result MDC. Result MDC Result MDC Result MDC 5271U1UV-01 2.97E+05 4.03E+04 3.77E+03 2.94E+04 4.41E+06 *2.83E+04 5.10E+05 2.57E+05 8.72E+04 3.51E+05 5271U1UV-02 3.97E+06 9.74E+04 2,61E+04 1.18E+05 5.18E+05 8.51E+04 3.40E+07 8.22E+05 2.24E+06 1.31E+06 5271U1UV-03 4.15E+06 1.16E+05 -1.96E+04 9.52E+04 4.11E+05 7:50E+04 2.73E+07 9.83E+05 1.71E+06 1.75E+06 5271 U1 UV-04 4.13E+06 1.08E+05 -1.84E+04 1.20E+05 7.08E+05 1.04E+05 3.78E+07 8.08E+05 1.90E+06 1.33E+06 5271 U1 UV-05 5.03E+06 1.49E+05 -4.41E+04 9.12E+04 8.66E+05 7.09E+04 2.94E+07 1.12E+06 1.82E+06 1.75E+06 5271U1UV-06 5.39E+05 3.89E+04 1.69E+04 3.59E+04 1.31E+06 3.24E+04 2.48E+06 3.52E+05 6.64E+05 8.03E+05 5271 U1 UV-07 2.53E+05 4.02E+04* -2.48E+03 2.25E+04 5.20E+06 2.93E+04 * -8.06E+04 6.81E+05 -4.47E+04 5.58E+05 5271 U1 UV-08 8.70E+05 6.54E+04 -2.48E+04 4.72E+04 1.60E+05 3.37E+04 4.37E+06 4.05E+05 1.56E+05 1.05E+06 5271U1UV-09 1.23E+05 9.04E+04 -1.61E+02 2.83E+04 4.29E+05 2.42E+04 6.64E+OS 1.71E+OS 8.0SE+04 4.06E+OS 5271 U1 UV-10Ja 2.10E+OS 5.23E+04 -2.18E+04 2.45E+04 1.40E+06 1.83E+04 5.57E+OS 4.54E+OS -4.82E+04 6.64E+05 5271U1UV-11J 1.24E+07 1.71E+OS -3.96E+04 1.06E+05 9.57E+07 1.08E+05 3.06E+07 2.27E+06 2.88E+06 3.26E+06 5271U1UV-12J 1.40E+07 1.55E+05 -7.00E+04 1.22E+05 3.01E+06 9.87E+04 4.63E+07 1.68E+06 2.60E+06 2.32E+06

J mdicates a Judgmental sample Zion Containment and Auxiliary Buildings B-1 5272-SR-03-0

Table B-2. ROC Concentrations in Unit 1 Undervessel Area Concrete Samples

. Gamma Spec Concentration (pCi/:m.2)a Sample ID ID Co-60 Cs-137 Cs-134 Eu-152 Eu-154 H-3 Ni-63 Sr-90 5271M0022 5271U1UV-11j 1.53E+06 3.76E+05 7.13E+03 1.74E+07 7.88E+05 2.61E+08 9.96E+05 4.66E+04 5271M0023 5271 U1 UV-05 4.62E+05 1.04E+04 2.66E+03 6.05E+06 2.04E+05 7.52E+07 3.58E+05 4.66E+04 5271M0024 5271 U1 UV-06 8.60E+04 1.22E+04 7.13E+02 9.78E+05 1.54E+04 1.80E+07 1.75E+05 -3.94E+04 5271M0025 5271 U1 UV-09 3.87E+04 1.11E+04 1.23E+03 4.97E+05

2.22E+04 1.29E+07 2.69E+05 3.58E+03 5271M0026 5271U1UV-03 1.45E+06 3.08E+05 ~3.89E+02 1.83E+07 6.20E+05 . 2.65E+08 2.81E+06 -1.43E+04 5271M0027 5271U1UV-04 . 5.54E+06 4.66E+04 2.59E+03 7.56E+07 3.13E+06 1.12E+09 6.48E+05 5.73E+04 5271M0028 5271U1UV-02 1.39E+06 3.94E+03 9.99E+03 1.87E+07 6.66E+05 2.77E+08 2.51E+05 -2.15E+04 5271M0029 5271U1UV-01 7.45E+04 4.62E+05 0.00E+OO 7.34E+05 7.16E+03 2.85E+07 1.49E+07 -3.58E+04 5271M0030 5271U1UV-07 8.95E+03 2.64E+05 0.00E+OO 1.q1E+04 -5.37E+03 2.72E+07 3.29E+05 0.00E+OO

Analytical results were converted from units ofpCi/g to pCi/m2 based on a concrete density of2.35 g/cm3 and sample depth of15.24 cm.

Zion Containment and Auxiliary Buildings B-2 5272-SR-03-0

. :Cq.::60*

Result *.MDC*' Result, :RisU:1t * :Result . Result 5271U1CMi-01 3.97E+04 3.14E+04

3.72E+03 1.66E+04 2.33E+OS 1.43E+04 7.01E+03 4.77E+04 -3.73E+04 6.57E+04 5271U1CM'T-02. 2.53E+04 1.12E+04

5.19E+02 2.17E+04 2.42E+OS 1.44E+04 -3.96E+04 4.19E+04 -1.07E+OS 6.57E+04 5271U1CMT-03 5.14E+04 3.17E+04 -5.59E+03 1.64E+04 1.37E+OS 1.66E+04 7.68E+03 4.25E+04 -6.10E+04 8.01E+04 5271U1CMT-04 3.97E+04 3.29E+04 1.90E+04 2.47E+04 6.57E+04 1.SOE+04 5.66E+04 3.86E+04 -2.71E+04 9.57E+04 5271U1CMT-05 2.12E+04 2.88E+04 -2.31E+03 1.70E+04 3.16E+OS 1.56E+04 2.17E+04 4.88E+04 -4.03E+03 6.99E+04 5271U1CMT-06 3.47E+04 9.28E+03 5.52E+03 1.80E+04 6.78E+04 1.22E+04 1.46E+04 4.46E+04 2.80E+04 7.87E+04 5271U1CMT-07 4.00E+04 9.30E+03 9.65E+03 1.84E+04 9.51E+04 1.23E+04 1.51E+03 4.52E+04 4.57E+04 7.51E+04 5271U1CMT-08 2.85E+04 3.01E+04 -1.18E+04 1.SSE+04 4.42E+04 1.27E+04 3.14E+04 4.26E+04 -2.49E+03 8.44E+04 5271U1CMT-09 1.96E+04 2.SOE+04 6.13E+02 1.33E+04 1.29E+04 2.0SE+04 -4.48E+03 3.72E+04 -1.99E+04 7.63E+04 5271U1CMT-10 4.26E+04 7.54E+03 -3.73E+03 1.73E+04 8.68E+04 1.70E+04 8.08E+01 4.68E+04 -3.39E+04 6.85E+04 Zion Containment and Auxiliary Buildings B-3 5272-SR-03-0 I ---

Table B-4. In Situ Gamma Spectrometry Measurements from Unit 2 Undervessel Area (pCi/m 2)

Measurement ID Co-60 Cs-134 Cs-137. Eu-152 Eu-154

Result .MDC Result MDC Result: .

MDC Result MDC Result MDC 5271U2UV-01 7.62E+03 1.93E+04 3.85E+03 1.78E+04 2.25E+04 1.11E+04 -2.62E+04 3.36E+04 -1.43E+04 6.46E+04 5271U2UV-02 5.92E+06 1.21E+05 -2.96E+04 1.21E+05 2.09E+06 1.30E+05 4.41E+07 1.08E+06 2.27E+06 1.71E+06 5271U2UV-03 5.95E+06 1.32E+05 -8.80E+04 1.06E+05 1.91E+06 8.54E+04 3.36E+07 1.15E+06 2.31E+06 2.20E+06 5271U2UV-04 4.99E+06 1.16E+05 6.34E+04 1.05E+05 2.07E+06 9.11E+04 3.23E+07 9.70E+05 1.75E+06 1.52E+06 5271 U2UV-05 1.07E+06 8.55E+04 -2.14E+04 3.97E+04 4.25E+05 3.77E+04 4.89E+06 5.46E+05 2.29E+05 1.29E+06 5271U2UV-06 9.54E+03 2.04E+04 6.35E+03 1.79E+04 8.56E+04 1.31E+04 2.07E+04 4.22E+04 -7.01E+04 5.89E+04 5271U2UV-07 1.06E+04 1.93E+04 1.71E+03 1.48E+04 5.00E+04 1.02E+04 -1.83E+04 3.67E+04 -3.28E+04 6.46E+04 5271U2UV-08 8.99E+06 1.58E+05 -2.16E+04 1.01E+05 5.04E+06 8.83E+04 3.01E+07 1.42E+06 1.92E+06 2.28E+06 5271U2UV-09 7.01E+05 7.08E+04 -1.42E+03 3.98E+04 6.21E+05 3.44E+04 3.32E+06 3.96E+05 9.81E+05 8.82E+05 Zion Containment and Auxiliary Bu'ildings B-4

5272-SR-03-0

Table B-5. ROC Concentrations in Unit 2 Undervessel Area Concrete Samples Gamma Spec Concentration (pCi/m2 a Sample ID ID Co~60 Cs-137 Cs-134 . . Eu-152 Eu-154 H-3 Ni-63 Sr-90 5271M0031 5271U2UV-9 1.01E+OS 7.45E+04 2.79E+03 1.35E+06 1.79E+04 3.78E+07 2.65E+OS -2.51E+04 5271M0032 5271U2UV-3 1.78E+06 2.61E+04 2.66E+03 2.0SE+07 1.0SE+06 2.69E+08 3.15E+05 -2.15E+04 5271M0033 5271U2UV-5 1.91E+OS 1.04E+04 -4.09E+02 2.06E+06 7.99E+04 5.91E+07 3.01E+OS 1.07E+04 5271M0034 5271U2UV-4 1.02E+06 4.69E+OS 5.84E+03 1.46E+07 5.66E+OS 2.01E+08 2.97E+OS -1.43E+04 5271M0035 5271U2UV-2 1.65E+OS 7.16E+03 5.84E+02 1.73E+06 5.48E+04 5.40E+07 2.01E+OS 5.01E+04 5271M0036 5271U2UV-8 9.8~E+OS 4.30E+04 2.59E+02 1.33E+07 4.80E+OS 1.53E+08 1.68E+OS 3.58E+04

Analytical results were converted from units of pCi/g to pCi/m2 based on a concrete density of 2.35 g/cm3 and sample depth of 15.24 cm.

Zion Containment and Auxiliary Buildings B-5 5272-SR-03-0

Table B-6. In Situ Gamma Spectrometry Measurements for Unit 2 565-ft Elevation and Above (pCi/m 2)

Co-60 Cs-134 C ~~ 137 E u,-152 .. ' *.* E u- 154

.Measurement-IP, " ..

Result .* MDC ' R,esqlt; :. MDC. Result. ****MDC .. .Re~idt .*

MPC Result * *MDC*

5271U2CMT-01 3.80E+04 9.90E+03 4.24E+03 1.90E+04 4.71E+04 1.18E+04 5.08E+03 3.91E+04 -3.90E+04 6.71E+04 5271U2CMT-02 2.26E+04 1.26E+04 -1.63E+03 1.73E+04 2.37E+04 9.05E+03 -1.14E+04 3.31E+04 2.79E+04 7.12E+04 5271U2CMT-03 1.48E+04 2.16E+04 -4.77E+03 1.33E+04 2.65E+04 2.17E+04 1.35E+04 3.91E+04 -7.43E+03 7.38E+04 5271 U2CMT~04 3.04E+04 6;7.5E+03 3.78E+03 1.66E+04 7.93E+04 1.29E+04 -4.84E+03 3.72E+04 -7.19E+03 6.85E+04 5271U2CMT-05 1.36E+04 2.29E+04 1.08E+04 1.84E+04 2.44E+04 2.59E+04 1.94E+04 4.12E+04 1.06E+04 6.85E+04 5271U2CMT-06 1.78E+04 2.54E+04 8.59E+03 1.90E+04 2.97E+04

1.13E+04 1.06E+04 3.87E+04 1.40E+04 7.38E+04 5271U2CMT-07 1.03E+04 2.29E+04 -1.11E+03 1.51E+04 3.05E+04 5.68E+03 -1.48E+04 3.64E+04 -1.12E+04 6.11E+04 5271 U2CMT-08 2.44E+04 2.58E+04 9.14E+03 1.77E+04 2.86E+04 2.36E+04. -1.22E+04 3.68E+04 -1.31E+04 6.57E+04 5271U2CMT~09 1.33E+04 2.25E+04 8.49E+03 1.77E+04 2.50E+04 9.33E+03 -1.66E+04 3.72E+04 -1.96E+04 6.71E+04 5271U2CMT-10 2.70E+04 2.73E+04 -2.62E+03 1.70E+04 3.04E+04 2.42E+04 -1.30E+04 4.29E+04 -4.13E+04 7.38E+04 Zion Containment and Auxiliary Buildings B-6 5272-SR-03-0

Table B-7. ROC Concentration in Auxiliary Building Concrete Samples (pCi/m 2t

. Sample ID Sample Type . Cs-137 Cs-134. . "

Sr-90 Co-60 1 .. : Ni-~3.

__ b 5271M0007 Random 2.12E+05 6.90E+03 8.31E+03 --

5271M0008 Random 4.75E+04 l.57E+03 9.87E+03 -- --

5271M0009 Random 4.84E+05 2.19E+03 l.09E+06 9.40E+o3* 4.75E+06 5271M0010 Random 3.73E+04 -l.88E+03 2.35E+03 -- --

5271M0011 Random 5.20E+04 7.05E+03 *3.13E+02 -- --

5271M0012 Random 8.04E+04 -l.57E+02 l.24E+04 -- --

5271M0013 Random l.65E+05 2.82E+03 l.08E+04 -- --

5271M0014 Random l.22E+04 2.82E+03 -2.04E+03 -- --

5271M0015 Random 9.36E+05 l.41E+03 7.52E+03 -3.60E+04 3.02E+05 5271M0016 Judgmental l.39E+09 3.29E+04 2.21E+07 5.74E+05 l.17E+09 5271M0017c Judgmental 9.32E+08 6.74E+03 1.44E+06 4.17E+05 4.47E+07 5271M0018 Judgmentai 1.51E+06 5.49E+03 2.15E+04 -- --

5271M0019 Judgmental 3.37E+06 -2.35E+03 4.65E+05 -- --

5271M0020 Judgmental 2,19E+07 2.35E+04 5.38E+06 4.07E+04 6.38E+07 5271M0021 Judgmental 8.82E+06 * -1.57E+03 l.38E+07 3.13E+04 2.37E+07

Analytical results were converted from units of pCi/ g to pCi/ m2 based on a concrete density of 2.35 g/ cm3 and sample depth of 6.7 cm.

bNot analyzed cCollected from the west wall of the Auxiliary Building Zion Containment and Auxiliary Buildings B-7 5271-SR-03-0

i I

I APPENDIX C: SURVEY AND ANALYTICAL PROCEDURES Zion Containment and Auxiliary Buildings 5272-SR-03-0

C.1 PROJECT HEALTH AND SAFETY I

l ' ' .

ORISE performed all survey activities in accordance with the ORA.CT Radiation Protection Manual, the ORA.CT Health and Sefety Manual, and the ORA.CT Radiological and Environmental Survry Procedures Manual (ORAU 2014, ORAU 2016c, and ORAU 2016a). 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 walkdq,wn 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 ORA.CT Radiological and Environmental Survry Procedures Manual (ORAU 2016a) or the project's 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 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 Radiological and Environmental Analytical Laboratory Procedures Manual (ORAU 2017)

ORAU Environmental Services and Radiation Training Quality Program Manual (ORAU 2016b)

The procedures contained in these manuals were developed to meet the requirements of U.S. Department of Energy (DOE) Order 414.lD and the NRC Qualify Program Manualfor the Office

<ifNuclear Material Sefety and Sefeguards and contain measures to assess processes during their performance.

Zion Containment and Auxiliary Buildings C-1 5271-SR-03-0

Quality control procedures include:

Daily instrument background and check-source measurements to confirm that equipment 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 IN SITU GAMMA SPECTROMETRY MEASUREMENTS

.Canberra's 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 Containment Buildings are discussed below. ORISE has previously reviewed ZS's 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.

C.3.1 Containment Buildings 565-ft Elevation and Above Radioactivity remaining in the 565-ft Elevation and Above of the Containment Buildings is surficial, as all of the contaminated concrete has been removed leaving the steel liner. A circular plane geometry was used to assess ROC concentrations in measurements collected from this area. The geometry template was modeled in the same manner as presented in ZS's TSD. The 0.635-cm-thick steel liner is assumed to corrode at a specified rate resulting in a corrosion. thickness of 0.153 cm that contains the residual contamination. Therefore, the circular plane model will consists of a steel source layer of thickness of 0.153 cm covering a non-radioactive steel backing of 0.482 cm.

Zion Containment and Auxiliary Buildings C-2 5271-SR-03-0

Figure C.1. !SOCS Circular Plan Template (Canberra 2009)

C.3.2 Undervessel Area C.3.2.1 Initial !SOCS Geometry Template For radionuclides that are present due to neutron activation- Eu-152/154 and Co-60-efficiency curves were generated using an exponential circular plane (ECP) template. The ECP template models con tamination as exponentially increasing to a maximum value and then exponentially decreasing as a function of depth. Figure C.2 illustrates the ECP and the necessary ISOCS inputs.

Zion Containment and .Auxiliary Buildings C-3 5271 -SR-03-0

C 0 - - - - Sc4.trca IAyfir 'IN"CINH (5..1) _ _ _ _ _ _....

8 C

0 0

J_

M _ ___.__........, 0.37 (ac) <'

oi--1 Depth Figure C.2. !SOCS Exponential Circular Plane Template (Canberra 2009)

In Figure C.2, the relaxation length (RL) (input 5.3) is the material depth where the relative activity concentration decreases by the value 1 / e (0.37) . The template input labeled as Dmax (5.2 in the figure above) is the depth at which the maximum activity concentration occurs . As much of the concrete in both Unit 1 and Unit 2 Undervessel Areas was removed during remediation, this parameter is not applicable. Therefore, the contamination depth profile for measurements in the Undervessel Area was described by two parameters, the relaxation length (5.3) and source layer thickness (5.1).

In order to determine the values of the two parameters that describe the contamination depth (inputs 5.2 and 5.3 described above), the ZS continuing characterization data set for each Containment Building was examined. Using the continuing characterization data sets, the 95% upper confidence level of the mean (UCL95) ROC concentration was determined for each 0.5 inch depth Zion Containment and Auxiliary Buildings C-4 5271-SR-03-0

increment in the concrete. The concentration depth profiles for each ROC in the Unit 1 Undervessel Area are provided in Figure C.3.

Co-60 Cs-137 I 60 . 10000 0) 40 20

\ ....

5000 0

3 C

0 0 0 0 5 10 15 20 0 5 10 15 20

p iC Eu-152 Eu-154 Q) 0 C 40

0 0

750

\

30 500 20 250 10 **

0 0 0 5 10 15 20 0 5 10 15 20 Depth (in)

Figure C.3. Unit 1 Concentration D epth Profiles for Gamma-emitting Radionuclides The concentration depth profile data were fitted to an exponential function of the form:

Where:

C(d) = concentration at depth d C 0

= concentration at depth d = 0 in d = depth in concrete RL = relaxation length.

2 Table C.1 provides the best-fit exponential equation for each ROC and the corresponding R value.

The R 2 value describes how well the variation in concentration is represented by the variation in Zion Containment and Auxiliary Buildings C-5 5271 -SR-03-0

depth by the exponential equation; an R2 value closer to 1.0 indicates a better fit. The depth of contamination was modeled as the depth in which the concentration was reduced to a fraction of 0.001. This depth is somewhat arbitrary as the relaxation length is relatively short such that the initial concentration is reduced to a fraction of approximately 0.02 after four RLs. Note that the curve fitting parameters below are different for each ROC; thus, ROC-specific energy curves were generated.

Table C.1. Unit 1 Undervessel Exponential Concentration Profile Parameters ROC RL (in) Co(pCi/g) (C(d)/Co) = 0.001 R2 Co-60 3.15 52 16 0.93 Cs-137 3.07 59 18 0.62 Eu-152 2.75 874 13 0.95 Eu-154 4.27 34 26 0.92 As indicated by the R 2 value in Table C.1, the exponential equation does not accurately represent the Cs-137 concentration depth profile, as Cs-13_7 was primarily a surface contaminant rather than a neutron activation product created within the concrete volume while the reactors were operational.

The CP template with uniform contamination serves as better model than the ECP template for Cs-137. The contamination depth was modeled as 0.5 in, which contains greater than 99% of the activity concentration.

The remaining parameters describing the ECP template are the source diameter, source density, and the source-to-detector distance. As with the CP, the source diameter is equal to two times the standoff distance. A nominal concrete density of 2.35 g/ cm3 was used for the source density.

The process outlined above was repeated for the Unit 2 Undervessel Area. Figure C.4 depicts the concentration depth profile for Unit 2. In some measurement locations all concrete was removed leaving only the steel liner. In these instances the same CP model as presented in Section C.3.1 was used. The curve fitting parameters are presented in Table C.2. As with Unit 1, the exponential curve does not accurately fit the Cs-137 concentration depth profiles; as such, the same CP template was applied to Unit 2 as in Unit 1.

Zion Containment and .Auxiliary Buildings C-6 5271-SR-03-0

Co-60 Cs-137 100 75 2000 50

- 1000 25 g

0 3 0 0 **** ********* * *** * ****** **** * ***

C 0 0 5 10 15 0 5 10 15 20

~

.b Eu-152 Eu-154 C

(1) 0 C

\

0 1500 60 0

1000 40 500 * *

- 20 0

0 0 5 10 15 20 0 5 10 Depth (in)

Figure C.4. Unit 2 Concentration Depth Profiles for Gamma-emitting Radionuclides Table C.2. Unit 2 Undervessel Exponential Concentration Profile Parameters ROC RL (in) Co(pCi/g) (C(d)/ Co) = 0.001 R2 Co-60 3.40 105 20.1 0.96 Cs-13 7 3.82 43 23 0.40 Eu-1 52 2. 73 1,681 16 0.97 Eu-154 4.13 65 24 0.96 C.3.22. Template Validation and Re.inement The Line Activity Consistency Evaluator (LACE), as part of the Genie 2000 software suite, was used to evaluate the ISOCS geometry templates. LACE provides a method to evaluate the consistency between the activities determined from different gamma ray energies of the same radionuclide.

E u-152 is a particularly useful radionuclide for LACE because of multiple gamma ray emissions across a wide energy range. Consistency is evaluated by plotting the ratio of the line activity and the weighted mean activity as a function of energy. A best-fit line is then determined based on the Zion Containment and Auxiliary Buildings C-7 5271 -SR-03-0

plotted data. If the !SOCS calibration template accurately represents the measurement geometry, the best fit line will have a slope of zero, indicating that all activities across* the line energy are consistent.

A positive slope indicates an underestimation of low-energy activity, meaning that not enough attenuation is accounted for in the low energy gamma rays. For the ECP template, a low estimate of the density or RL would result in an overestimate of the low-energy efficiency resulting in a lower low-energy line activity and positive slope. A negative slope indicates the opposite. An example LACE output for Eu-152 using the initial ECP template is shown in Figure C.5.

EU-152: Weighted Mean Slope= 0.072(+ -0 .034) 2

1Q -,-~~~~~~~~~~~~~~~~~~~~~~~~~~

0 9*1Q-1 2*102 4*1 0 2 Energy (kEv)

Figure C.5. Example LACE Output for Eu-152 Using the Initial ECP Template Detector efficiency is most sensitive to the relaxation length and density parameters in the ECP template; the efficiency is also sensitive to the sourceto-detector distance and the source diameter; however, these parameters are relatively well understood. Separately varying the RL from 2.73 to 8.0 and density from 2.35 to 6.40 g/ cm3 resulted in a slope of approximately zero for Eu-152 as indicated by the LACE outputs. The !SOCS uncertainty estimator was used to assess the relative magnitude of change in efficiency due to varying the RL and source density. The change in efficiency is summarized in Table C.3. Note that the values in Table C.3 are positive, indicating a positive increase.

Zion Containment and Auxiliary Buildings C-8 5271 -SR-03-0

Table C.3. Fractional Change in Efficiency for ECP Template Parameters Parameter Co-60 Cs-134 Cs-137 Eu-152 Eu-154 Density (2.35 to 6.40 g/ cm3) 0.7548 0.7362 0.7314 0. 7154 0.7339 RL (2. 73 to 8.00 in) 0. 7736 0. 7558 0. 7512 0. 7351 0.7 536 As indicated by Table C.3, varying the RL and source density to achieve an acceptable LACE output results in the same change in efficiency. Of both parameters, the source density is the least understood, as the bulk density of the source is heavily influenced by rebar present in the concrete.

Therefore, the source density at each measurement location was varied within the bounds presented in Table C.3 to minimize the LACE slope. An example LACE output after optimizing the source density is presented in Figure C.5.

Weighted Mean Slope= -0.000(+ -0 .034)

EU- 152 :*

2*1 0 0 10° I

f 9*10-1 -- I I I I I I I I I I I I I I 2*1 0 2 4*10 2 6* 102 8*1 0 2103 Energy (ke Figure C.5. Example LACE Output After Optimizing the Source D ensity C.4 RADIOLOGICAL SAMPLE ANALYSIS C.4.1 Gamma Spectroscopy Samples were analyzed as received, mixed, crushed, and/ or homogenized as necessary, and a portion sealed in a 0.5-liter Marinelli beaker. The quantity placed in the beaker was chosen to reproduce the Zion Containment and Auxiliary Buildings C-9 5271 -SR-03-0

calibrated counting geometry. Net material weights were determined and the samples counted using

I intrinsic, high purity, germanium detectors coupled to a pulse height analyzer system. Background I

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 radionuclides of concern were reviewed for consistency of activity.

Spectra were also reviewed for other identifiable TAPs. TAPs used for determining the activities of the radionuclides of concern and the typical associated MDCs for a one-hour count time are presented in Table C.4.

Table C.4. Typical MDCs Total Absorption Peak Radionuclide TAP (MeV)a MDC (p_Ci/ g) *

~o-60 1.332 0.06 Cs-134 0.795 0.05 Cs-137 0.662 0.05 Eu-152 0.344 0.10 Eu-154 0.723 0.15 a mega electron volt C.4.2 Ni-63 Analysis Soil 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 removec;l 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 60-minute count time using this procedure is 0.8 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 is coprecipitated on lead sulfate. 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 Zion Containment and Auxiliary Buildings C-10 5271-SR-03-0

counted on a low-background gas proportional counter. The typical MDC for a 60-minute count time using this procedure is 0.4 pCi/ g.

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 ina stream of oxygen gas and passes the products (including carbon dioxide and H 20 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.

C.4.5 Detection Limits Detection limits, referred to as MDCs, were based on 95% confidence level. Because of variations in background levels, measurement efficiencies, and contributions from other radionuclides in samples, the detection limits differ from sample to sample and instrument to instrument.

.i Zion Containment and Auxiliary Buildings C-11 5271-SR-03-0

APPENDIX D: MAJOR INSTRUMENTATION Zion Containment and Auxiliary Buildings 5271-SR-03-0

T 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 Nal 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, Broa~ 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 SO mm, 90-degree POV 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 Containment and Auxiliary Buildings D-1 5271-SR-03-0

(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-l\1X5970-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 Proportion?il 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)

Zion Containment and Auxiliary Buildings D-2 5271-SR-03-0

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