ML16006A200

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Final Report - Confirmatory Survey Activities Associated with Phase 2 for the Buffalo Materials Research Center at the State University of New York, University at Buffalo, Buffalo, New York
ML16006A200
Person / Time
Site: University of Buffalo
Issue date: 01/06/2016
From: Harpenau E
Oak Ridge Associated Universities
To: Tanya Smith
Reactor Decommissioning Branch
John Clements, 301-415-5878
References
RFTA 12-010 DCN: 5177-SR-02-1
Download: ML16006A200 (56)
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FINAL REPORTCONFIRMATORY SURVEY ACTIVITIES ASSOCIATED WITH PHASE 2 FOR THE BUFFALO MATERIALS RESEARCH CENTER AT THE STATE UNIVERSITY OF NEW YORK, UNIVERSITY AT BUFFALO, BUFFALO, NEW YORK EVAN M. HARPENAU Prepared for the U.S. Nuclear Regulatory Commission January 2016 Further dissemination authorized to the NRC only; other requests shall be approved by the originating facility or higher NRC programmatic authority.

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

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

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

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

FINAL REPORTCONFIRMATORY SURVEY ACTIVITIES ASSOCIATED WITH PHASE 2 FOR THE BUFFALO MATERIALS RESEARCH CENTER AT THE STATE UNIVERSITY OF NEW YORK, UNIVERSITY AT BUFFALO, BUFFALO, NEW YORK FINAL REPORT Prepared by Evan M. Harpenau JANUARY 2016 Prepared for the U.S. Nuclear Regulatory Commission Prepared by ORAU under the Oak Ridge Institute for Science and Education contract, number DE-AC05-06OR23100, with the U.S. Department of Energy under interagency agreement (NRC FIN No. F-1244) between the U.S. Nuclear Regulatory Commission and the U.S. Department of Energy.

BMRC-Phase 2 Confirmatory Survey Report 5177-SR-02-1

CONTENTS TABLES ......................................................................................................................................................... ii FIGURES.. ......................................................................................................................................................... ii ACRONYMS .................................................................................................................................................... iii EXECUTIVE

SUMMARY

............................................................................................................................. iv

1. INTRODUCTION....................................................................................................................................... 1
2. SITE DESCRIPTION ................................................................................................................................. 2
3. OBJECTIVES................................................................................................................................................ 2
4. RADIONUCLIDES OF CONCERN ...................................................................................................... 2
5. PROCEDURES ............................................................................................................................................ 5 5.1 DOCUMENT REVIEW ..................................................................................................................... 7 5.2 REFERENCE SYSTEM ..................................................................................................................... 7 5.3 SURFACE SCANS ............................................................................................................................. 7 5.4 SURFACE ACTIVITY AND GAMMA MEASUREMENTS ................................................................. 8 5.5 SOIL AND REMOVABLE ACTIVITY (SMEAR) SAMPLING ............................................................ 8 5.6 SPLIT SOIL SAMPLE COMPARISON ............................................................................................... 9
6. SAMPLE ANALYSIS AND DATA INTERPRETATION ...............................................................10
7. FINDINGS AND RESULTS ...................................................................................................................10 7.1 DOCUMENT REVIEW ...................................................................................................................10 7.2 IN-PROCESS OBSERVATIONS......................................................................................................11 7.3 SURFACE SCANS ...........................................................................................................................11 7.4 SURFACE ACTIVITY MEASUREMENTS .......................................................................................12 7.5 RADIONUCLIDE CONCENTRATIONS IN BEDROCK AND SOIL ...............................................13
8. COMPARISON OF RESULTS WITH GUIDELINES .....................................................................13 8.1 COMPARISON OF SURFACE ACTIVITY MEASUREMENTS.........................................................13 8.2 ANALYTICAL COMPARISON OF SAMPLES..................................................................................13
9.

SUMMARY

..................................................................................................................................................14

10. REFERENCES .........................................................................................................................................15 BMRC-Phase 2 Confirmatory Survey Report i 5177-SR-02-1

TABLES Table 4.1. BMRC Radionuclides of Concern................................................................................................. 3 Table 4.2. NRC License Termination Screening Levels for Building (Bedrock) Surfaces ...................... 4 Table 4.3. DCGLs for Primary Radionuclides of Concern in Soil ............................................................. 4 Table 5.1. Final Status Survey Unit and Confirmatory Unit Crosswalk .................................................... 6 Table 5.2. Confirmatory Unit Sampling Frequency ...................................................................................... 9 Table B-1. Surface Activity Levels ..............................................................................................................B-1 Table B-2. Radionuclide Concentrations in Soils ......................................................................................B-4 Table B-3. Radionuclide Concentrations for Judgmental Bedrock and Soil Samples .........................B-5 Table B-4. Side-by-Side Measurements for Survey Unit 4.......................................................................B-6 Table B-5. Side-by-Side Measurements for Survey Unit 6.......................................................................B-7 Table B-6. Summary Statistics for Split Soil Samples ...............................................................................B-8 FIGURES Figure 5.1 Survey Unit Classification .............................................................................................................. 6 Figure A-1. Location of Buffalo Materials Research Center, Buffalo, New York .............................. A-1 Figure A-2. Confirmatory Units and Designated Classifications ........................................................... A-2 Figure A-3. Gamma Walkover Survey of Confirmatory Unit 1............................................................. A-3 Figure A-4. Gamma Walkover Survey of Confirmatory Unit 2............................................................. A-4 Figure A-5. Gamma Walkover Survey of Backfill Lift 1 and Adjacent Stockpile Footprint Area ... A-5 Figure A-6. Gamma Walkover Survey of Backfill Lift 2 ......................................................................... A-6 Figure A-7. Gamma Walkover Survey of Stockpile Area ....................................................................... A-7 Figure A-8. Planned Random Sample Locations in Confirmatory Units 1 and 2 ............................... A-8 Figure A-9. Actual Random Sample Locations in Confirmatory Units 1 and 2 .................................. A-9 Figure A-10. Judgmental Sample Locations in Confirmatory Units 1 and 2 ..................................... A-10 Figure A-11. Side-by-Side Smear/Measurement Locations in Survey Unit 4.................................... A-11 Figure A-12. Side-by-Side Measurement Locations in Survey Unit 6 ................................................. A-12 BMRC-Phase 2 Confirmatory Survey Report ii 5177-SR-02-1

ACRONYMS BMRC Buffalo Materials Research Center CU confirmatory unit D&D decontamination and decommissioning DCGL derived concentration guideline level DP decommissioning plan FSS final status survey FSSP final status survey plan GPS global positioning system GWS gamma walkover survey INEEL Idaho National Engineering and Environmental Laboratory MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDC minimum detectable concentration NaI sodium iodide NIST National Institute of Standards and Technology NRC U.S. Nuclear Regulatory Commission ORISE Oak Ridge Institute for Science and Education PSP project-specific plan REAL Radiological and Environmental Analytical Laboratory ROC radionuclide of concern SOF sum-of-fractions SU survey unit SUNY State University of New York UB University at Buffalo BMRC-Phase 2 Confirmatory Survey Report iii 5177-SR-02-1

FINAL REPORTCONFIRMATORY SURVEY ACTIVITIES ASSOCIATED WITH PHASE 2 FOR THE BUFFALO MATERIALS RESEARCH CENTER AT THE STATE UNIVERSITY OF NEW YORK, UNIVERSITY AT BUFFALO, BUFFALO, NEW YORK EXECUTIVE

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) requested that ORAU, working under the Oak Ridge Institute for Science and Education (ORISE) contract, perform independent confirmatory surveys at the State University of New York, University at Buffalo (UB) in Buffalo, New York. Enercon, the licensees contractor, and UB are nearing completion of decontamination and decommissioning activities of the Buffalo Materials Research Center (BMRC) with the goal of satisfying the NRC-approved license termination screening levels for both building (bedrock) surfaces and impacted soils. This report focuses on confirmatory survey activities performed in support of the BMRC excavation, surrounding soils, backfill material, and soil laydown areas.

ORAU performed independent assessment activities including gamma and beta radiation surveys, along with bedrock and soil sampling during the periods of January 26-29, February 3-6, and August 17-21, 2015. Confirmatory survey activities included surveys of six final status survey units, the soil reuse piles, and portions of the laydown area for the reuse piles. Additionally, ORAU performed 27 side-by-side direct measurements; 32 independent direct measurements with either smears or volumetric samples; and 4 split soil samples. The results of ORAU gamma and beta radiation surveys, combined with laboratory analytical results from smears and bedrock and soil samples, support the conclusion that the BMRC and impacted land areas associated with it satisfy the approved license termination screening levels described in the licensees decommissioning and final status survey plans.

BMRC-Phase 2 Confirmatory Survey Report iv 5177-SR-02-1

FINAL REPORTCONFIRMATORY SURVEY ACTIVITIES ASSOCIATED WITH PHASE 2 FOR THE BUFFALO MATERIALS RESEARCH CENTER AT THE STATE UNIVERSITY OF NEW YORK, UNIVERSITY AT BUFFALO, BUFFALO, NEW YORK

1. INTRODUCTION The U.S. Nuclear Regulatory Commission (NRC) requested that ORAU, via the Oak Ridge Institute for Science and Education (ORISE) contract, perform confirmatory survey activities of the Phase 2 decontamination and decommissioning (D&D) activities for the Buffalo Materials Research Center (BMRC) which is owned by the State University of New York (SUNY) at the University at Buffalo (UB). Designed and constructed between 1959 and 1961 by American Machine and Foundry Atomics, the BMRC was a research and test reactor facility with a pool-type reactor. The initial criticality date for the reactor was March 24, 1961 and the last day of operation was June 23, 1994.

Since June 6, 1997, the facility has been in a possession-only licensing status and the unused fuel was shipped to North Carolina State University in 1998. The spent fuel was shipped to Idaho National Engineering and Environmental Laboratory (INEEL) in 2005. Because there is no future need for the BMRC, the facility is being decommissioned (Enercon 2012a).

During its operating history, the BMRC was used for training and education, transient fuel performance testing, nuclear component testing and calibration, materials radiation damage research, isotope production, and neutron interrogation through activation analysis, radiography, and delayed fission assay. The licensees contractor, Enercon, conducted a historical site assessment and a characterization survey to assess and detail the radiological status of the BMRC (Enercon 2010 and 2011).

Currently, the site is in the process of being decommissioned. Enercon prepared a decommissioning plan (DP) and a final status survey plan (FSSP) (Enercon 2012a and 2012b); both were approved by the NRC. The DP provides guidance on the processes and methods to be used to safely decontaminate, remove, and dispose of radioactive materials, equipment, systems, components, and soil associated with the BMRC. The FSSP details the survey and sampling methodologies Enercon BMRC-Phase 2 Confirmatory Survey Report 1 5177-SR-02-1

would implement to demonstrate compliance with the unrestricted release criteria for the BMRC, and receive termination of NRC license R-77.

2. SITE DESCRIPTION The UB campus is approximately 20 miles south of the Canadian border, 5 miles east of the Niagara River, 90 miles west of Rochester, and 80 miles from the southern border of New York with Pennsylvania, in Buffalo, Erie County, New York. The BMRC is located on the southern edge of the South Campus of the UB off of Rotary Drive (Figure A-1).

The BMRC facility consisted of the reactor, the Containment Building which enclosed the reactor and other ancillary facilities related to the use of the reactor, and the Administrative Building (also referred to as the Laboratory Wing) which contained offices, classrooms, and laboratories. The structures associated with the BMRC have been demolished and removed from the site. At the time of confirmatory survey activities, the site consisted of exposed bedrock where the BMRC facility was located, and the impacted soils surrounding the excavation.

3. OBJECTIVES The objectives of the confirmatory survey activities were to provide independent contractor field data reviews and to generate independent radiological data for use by the NRC in evaluating the accuracy and adequacy of the licensees procedures and final status survey (FSS) results.
4. RADIONUCLIDES OF CONCERN The primary radionuclides of concern (ROCs) for the BMRC are beta-gamma emittersfission and activation productsresulting from reactor operation. Table 4.1 provides a comprehensive list of the ROCs for the BMRC and the areas of concern/matrix.

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Table 4.1. BMRC Radionuclides of Concerna Half-life Radionuclide Emission Area(s) of Concern (years)

Ag-108mb 438 , Soil; tank water; SSCsc Am-241 432 , Tank sediment C-14 5,730 Laboratory areas Co-60 5.27 , Soil; SSCs; Bioshield Cs-137 30.1 ,d Soil; SSCs; Bioshield Eu-152 13.6 , Soil; SSCs; Bioshield Eu-154 8.59 , Soil; SSCs; Bioshield H-3 12.3 Soil; Bioshield; tank water Ni-63 100 Soil; SSCs; Bioshield Pu-238 87.8 , Tank sediment Pu-239 24,100 , Tank sediment Pu-240 6,600 , Tank sediment Sr-90 28.8 Soil; SSCs; ventilation systems aTable source: Enercon 2012b.

bBold text identifies ROCs addressed during confirmatory surveys as requested by the NRC.

cSSCs = structures, systems, and components d emission from Ba-137m progeny As previously stated, the remaining impacted areas consisted of an excavation with exposed bedrock and the impacted soils surrounding it. As a result of the two material types, confirmatory survey results were compared to two different sets of release criteria. First, the surface activity screening values implemented for bedrock surfaces were consistent with Appendix B of NUREG-1757, Consolidated Decommissioning Guidance. Those values are represented in Table 4.2 with the understanding that Co-60 represents the limiting ROC. Note, however, that a screening level for Ag-108m is not provided in NUREG-1757; thus the Table 4.2 value was derived and presented in the DP.

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Table 4.2. NRC License Termination Screening Levels for Building (Bedrock) Surfaces Surface Screening Levelsa Radionuclide (dpm/100cm2)

Ag-108m 17,000 C-14 3,700,000 Co-60 7,100 Cs-137 28,000 H-3 120,000,000 Ni-63 1,800,000 Sr-90 8,700 aScreening levels are based on the assumption that the fraction of removable surface contamination is equal to 0.1. For cases when the fraction of removable contamination is undetermined or higher than 0.1, users may assume, for screening purposes, that 100 percent of surface contamination is removable, and therefore the screening levels should be decreased by a factor of 10. All values are from NUREG-1757, except for Ag-108m which is a derived, site-specific guideline.

Secondly, the impacted soils were compared to the volumetric derived concentration guideline levels (DCGLs) provided in Table 3-1 of the FSSP. Table 4.3 reproduces these DCGLs.

Table 4.3. DCGLs for Primary Radionuclides of Concern in Soila NRC Screening Value for Selected DCGL Value Radionuclide Surface Soils (pCi/g) (pCi/g)

Ag-108m None 8.2 Am-241 None 2.1 C-14 12 12 Co-60 3.8 3.8 Cs-137 11 11 Eu-152 8.7 6.9 Eu-154 8 8 H-3 110 110 Ni-63 2,100 2,100 Pu-238 2.5 2.5 Pu-239/240 2.3 2.3 Sr-90 1.7 1.7 aFrom Table 3-1 of the BMRC final status survey plan (Enercon 2012b)

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Each radionuclide-specific soil DCGL represents the concentration above background of a residual radionuclide that would result in a radiological dose of 25 millirem per year (mrem/yr) to the average member of the critical group, except for Eu-152, which was selected below the 25 mrem/yr limit for administrative reasons. For consistency with the licensees application of DCGLs to gross soil concentrations such that data were not corrected for background contributions, ORAU also reported analytical results without background corrections. Because multiple contaminants could be present, the unity rule must also be applied to each sample to ensure compliance with the dose limit.

Compliance with the unity rule is based on the sum-of-fractions (SOF) calculation as follows:

Cj SOFTOTAL = SOFj =

DCGL,j

=0

=0 Where Cj is the concentration of ROC j, and DCGL,j is the DCGL for ROC j. Note that gross concentrations are considered here for conservatism.

The unity rule is satisfied when the SOF does not exceed the value of 1.

5. PROCEDURES During the periods of January 26-29, February 3-6, and August 17-21, 2015, ORAU personnel conducted document reviews, independent radiation surface activity measurements, and volumetric sampling to evaluate the radiological status of the exposed soil/bedrock, surrounding areas, and backfill soils at the location of the former BMRC. The area associated with the BMRC was divided into seven survey units (SUs) that were classified per Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) guidance (NRC 2000). The excavation was divided into four Class 1 units, two Class 2 units comprised of side slope and surrounding soils, and a Class 3 unit encompassing the remaining project area. The FSS units were combined into confirmatory units (CUs) in accordance with the confirmatory project-specific plan (PSP) (ORAU 2014) (Figure A-2). Figure 5.1 depicts the SUs implemented for FSS by classification and description. Table 5.1 provides the crosswalk between the FSS units and the corresponding CUs along with the confirmatory survey specifications, respectively. Survey activities performed by ORAU personnel were conducted in accordance with the NRC-approved project-specific plans, the ORAU Radiological and Environmental BMRC-Phase 2 Confirmatory Survey Report 5 5177-SR-02-1

Survey Procedures Manual, and the ORAU Environmental Services and Radiation Training Quality Program Manual (ORAU 2013, 2014, 2015a, and 2015b).

Survey Description Unit 1 N16 tank area 2 Tank farm area 3 Sub-basement 4 Containment 5 Side slopes Cooling tower, 6 sewer, tank farm plumbing 7 Remaining area Figure 5.1 Survey Unit Classification Table 5.1. Final Status Survey Unit and Confirmatory Unit Crosswalk Specificationsa Planned Activities Walkover Random/Start Total Area (Composite) Judgmental CU FSS Designation (m2) Type Coverage Samples Samples SU 1, SU 2, SU 3, 1 915 High density 15(3) 0-1 and SU 4 Medium 2 SU 5 and SU 6 1,159 10(2) 0-1 density aSpecifications were derived from Table 3-7 of the Revised Final Status Survey Plan and the BMRC Stockpile Sampling Letter (Enercon 2012b and 2014).

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The limiting conditions encountered onsite (e.g., frozen bedrock, deep snow cover, and heavy rain) prevented ORAU from performing confirmatory survey activities in exact accordance with the PSP.

With NRC approval, the survey and sampling approach described in the PSP was modified to account for those limiting conditions. The specific necessary modifications are detailed in the following subsections.

5.1 DOCUMENT REVIEW Prior to on-site activities, ORAU reviewed the DP, FSSP, and FSS packages. Each document and associated data were reviewed for adequacy and appropriateness while taking into account MARSSIM guidance (NRC 2000). While on site, ORAU staff were able to perform preliminary reviews of field documents associated with the Class 1 FSS units.

5.2 REFERENCE SYSTEM ORAU referenced survey results using Enercons established coordinate system. The reference system used global positioning system (GPS) coordinates X (easting) and Y (northing), based on New York State Plane North American Datum 1983 coordinates.

5.3 SURFACE SCANS Surface scans for gamma radiation were performed using Ludlum Model 44-10 sodium iodide (NaI) scintillation detectors coupled to ratemeter-scalers with audible indicators. The detectors were coupled to GPS equipment that enabled real-time gamma count rate and spatial data capture.

Locations of elevated direct radiation, suggesting the presence of residual contamination, were recorded and identified for further investigation. High-density scans were performed in the CU 1 excavation, and medium-density scans were performed on the side slopes in CU 2 (Figures A-3 and A-4). High-density scans were also performed of two compacted backfill lifts and two stockpile location footprints (Figures A-5 through A-7).

Surface scans of exposed bedrock for beta radiation were also planned. However, the beta scans discussed in the confirmatory survey plan were not performed during the two winter 2015 confirmatory trips due to the extensive snow cover and limited accessibility to the bedrock surface.

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Only gamma radiation scans of soils were required for the walkover surveys during the August 2015 trip.

5.4 SURFACE ACTIVITY AND GAMMA MEASUREMENTS Direct beta surface activity measurements were collected on snow-cleared bedrock to compare measured results to the screening values identified in Table 4.2. Measurements were performed using Ludlum Model 44-142 scintillation detectors coupled to ratemeter-scalers at 59 locations that were selected either through random-start systematic or judgmental processes. Qualitative gamma measurements were also performed at each beta measurement location to assess the potential for the presence of elevated concentrations of gamma emitting ROCs. Gamma measurements were performed using Ludlum Model 44-10 NaI scintillation detectors coupled to ratemeter-scalers.

5.5 SOIL AND REMOVABLE ACTIVITY (SMEAR) SAMPLING ORAU originally planned the confirmatory activities where soil samples were to be collected from each SU and the data used to estimate the mean contaminant concentrations. The planning outputs resulted in the intent to collect three 5-increment composite samples from CU 1 and two 5-increment composite samples from CU 2. The respective increment locations were distributed in a random or systematic fashion within the respective CUs as shown on Figure A-8.

However, prior to field activities, heavy snow and ice buildup on flat surfaces adversely impacted the ability to collect volumetric samples from the bedrock surface of CU 1. Therefore, the composite sampling approach was replaced, with NRC concurrence, with direct total and removable beta activity measurements at the random-start systematic locations that corresponded with the CU 1 bedrock surfaces. Two of these random/systematic locations were inaccessible due to the presence of the soil ramp being used to access the excavation.

Within CU 2 the planned composite increment locations were instead collected as independent samples for analysis rather than combining into composites. This modification was also approved by NRC prior to initiating CU 2 surveys. Figure A-9 shows the location of individual direct measurements/smear sample location in CU 1noting that locations 5177R0034 and 5177R0035 were not collected due to inaccessibilityand soil sample locations in CU 2.

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In addition to the above random/systematic samples, the NRC requested that judgmental miscellaneous samples be collected from each layer of bedrock observed in the sump areas within the trench of CU 1. As a result, six judgmental samples were collected from the two sump areas (numbered 5177M0001-5177M0006). Judgmental soil samples were also collected from the two FSS locations with the highest gamma radiation levels in SU 6. (numbered 5177M0007 and 5177M0008).

Though these two samples were comprised of soil, the judgmental sampling number sequence was used to maintain consistent identification of judgmental locations. Smears were also collected at FSS bedrock locations in SU 4. Figure A-10 provides the locations of the judgmental samples collected in SUs 4 and 6 (numbered 5177M0001-5177M0008). The planned versus final sampling frequency is summarized in Table 5.2.

Table 5.2. Confirmatory Unit Sampling Frequency Random Sampling Judgmental Sampling Description/ Number of CU Number of Number of Class Number of Judgmental Random Judgmental Samples Samples Samples Planned Samples Planned Collecteda Collecteda 1 Class 1 excavation 3 (15 increments) 0b 0-1 6 Class 2 slopes and 2 2 (10 increments) 10 0-1 2 surrounding soils aLocations for confirmatory samples collected within their respective CUs are provided in Appendix A.

bDirect beta measurements and smears were collected on bedrock per NRC direction.

Smears for determining removable gross alpha/beta activity levels were also collected from 10 side-by-side FSS measurements locations in SU 4. Figure A-11 shows the location of smears collected in SU 4 (numbered 5177R0012-R0021). Identification numbers for bedrock measurement locations are 5177R0012-5177R0036 (smears 5177R0034 and 5177R0035 were not collected/submitted). Side-by-side measurements were also made in SU 6. Smears were not collected at these locations because there was no exposed bedrock (Figure A-12).

5.6 SPLIT SOIL SAMPLE COMPARISON Finally, NRC requested that ORAU analyze split samples from each of the stockpile soils for comparison with the corresponding Enercon results. Enercon had retained four such samples that equally represented grab samples from the Class 1 and Class 2 stockpiles. The ORAU identification numbers for the split samples are 5177S0021-5177S0024.

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6. SAMPLE ANALYSIS AND DATA INTERPRETATION Samples and data collected on site were delivered to the ORAU/ORISE facility for analysis and interpretation. Sample custody was transferred to the ORAU/ORISE Radiological and Environmental Analytical Laboratory (REAL) in Oak Ridge, Tennessee. Sample analyses were performed in accordance with the ORAU Radiological and Environmental Analytical Laboratory Procedures Manual (ORAU 2015c). Soil samples were analyzed by gamma spectroscopy for gamma-emitting ROCs. Samples from the Class 1 excavation were also analyzed for Sr-90. Soil sample analytical results were reported in units of picocuries per gram (pCi/g). Smear samples collected for the quantification of gross alpha/beta activity were analyzed using a low-background proportional counter. Smear sample and direct measurement results are reported in units of disintegrations per minute per one hundred square centimeters (dpm/100 cm2).
7. FINDINGS AND RESULTS The results of the confirmatory survey are discussed in the subsections below.

7.1 DOCUMENT REVIEW ORAU conducted a high-level review of the licensees DP and FSSP in conjunction with a detailed review of the Class 1 FSS survey package (Enercon 2012a, 2012b, 2015). The reviews indicated that the general procedures and methods implemented were appropriate for meeting the release criteria.

However, one minor issue related to the performance of gamma scan surveys was identified during the FSS package review and was submitted to NRC via email correspondence (ORAU 2015d).

Specifically, package instructions did not clearly state whether FSS gamma surveys would be performed prior to moving residual soils or after bedrock was exposed.

Reviews of additional FSS field documentation and electronic data entry tables were performed as part of on-site confirmatory survey activities. An observation was made, and communicated to the NRC inspector, that the contractor was not recording all critical survey information in the field documents nor as electronic data. Specifically, the model number for the ratemeter used during the survey was the only instrument-related information on field data sheets. Without a direct connection BMRC-Phase 2 Confirmatory Survey Report 10 5177-SR-02-1

to detector-specific information on the field sheets, ORAU was not able to confirm that all of the appropriate detector inputs were used in the surface activity calculations.

Post-survey reviews of the contractors direct measurement data identified two data quality issues.

The first corresponded to an improper detector surface area being applied to the surface activity calculations that led to an under-reporting of direct radiation levels. Secondly, all of the SU 3 bedrock FSS surface activity data were negative values. This was indicative of applying inappropriate reference area background count rates to the gross measurement results which also led to net direct radiation levels being under-reported. Both issues were presented to the NRC who then discussed with the licensee.

7.2 IN-PROCESS OBSERVATIONS In-process observations identified two issues that were inconsistent with the methods described in the FSSP and standard industry practices. First, the detector stand-off during gamma walkover surveys (GWSs) in the excavation was in excess of 12 inches above the surface. Standard industry practice and scan minimum detectable concentration determinations are in most cases based on maintaining the detector at a nominal distance from the surface. The second issue involved a deviation to the sequence of survey activities called out in the FSS package. The FSS sequence was to first conduct the GWS, followed by the FSS sampling which would be in accordance with standard industry practice and MARSSIM guidance. However, the contractor was observed collecting direct measurements and samples prior to performing the GWS in the excavation and of the surrounding soils. ORAU conveyed this information to the NRC.

7.3 SURFACE SCANS A majority of the confirmatory gamma scan results from the base of CU1 exhibited radiation levels within the detector background range of 3,100-7,000 cpm. Gamma radiation count rates increased up to 9,000 cpm as scans approached the walls of the excavation (Figure A-3). The increased count rate was expected due to the additional excavation wall soil surface area geometry. Scans did identify two isolated areas for further investigation where the count rate was in excess of 9,000 cpm. Only the location in the eastern-most corner of SU 3, where the count rate exceeded 19,000 cpm, was BMRC-Phase 2 Confirmatory Survey Report 11 5177-SR-02-1

marked for further evaluation once the tarpsplaced to prevent sloughing of soil onto the bedrockwere removed.

Gamma scans of the CU 2 slopes and surrounding soils after the contractor removed the tarps also exhibited radiation levels near background, though the data suggested the presence of two distinct background populations. Scans of the slopes indicated background levels ranging from 7,400-11,000 cpm, while the radiation levels in the surrounding soils ranged from 4,100-9,600 cpm (Figure A-4).

The same location of elevated direct gamma radiation in SU 3 was observed while scanning the lower slope in the corner of SU 5. However, with the tarps removed, the count rate for that location increased to approximately 37,000 cpm. The NRC was notified and the contractor began further investigations of the contaminated area.

Similar to the majority of surface scan results, confirmatory gamma scans of the two compacted backfill lifts, and the two soil stockpile footprints were consistent with background or did not warrant further investigation. Results ranged from approximately 7,000-11,800 cpm (Figures A-5 to A-7), noting the background (without geometric effects) of about 8,000 cpm.

7.4 SURFACE ACTIVITY MEASUREMENTS The majority of the direct beta radiation activity results are consistent with typical background levels.

Eight of the confirmatory measurement locations exhibited beta activity above the detectors minimum detectable concentration (MDC). The highest direct beta confirmatory surface activity measurement was 1,700 dpm/100 cm2 at FSS locations SU6-05, SU6-11 and Judgmental 9 (in SU6).

Surface activity for the remaining confirmatory measurement locations ranged from -1,800 to 1,500 dpm/100 cm2. Table B-1 provides a summary of the confirmatory direct measurement data for each location addressed during the survey.

No removable activity was detected; results ranged from -1 to 3 dpm/100 cm2 alpha and -4 to 5 dpm/100 cm2 beta with MDCs of 12 dpm/smear and 15 dpm/smear, respectively.

Location-specific removable activity results are summarized in Table B-1.

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7.5 RADIONUCLIDE CONCENTRATIONS IN BEDROCK AND SOIL All confirmatory soil and five judgmental bedrock samples exhibited ROC concentrations below the respective MDCs for gamma spectroscopy and Sr-90 analysis (Tables B-2 and B-3). Judgmental sample 5177M0003 was the only sample that contained a ROC, Cs-137, concentration above the analytical MDC. However, the Cs-137 concentration of 0.12 pCi/g only represents a small fraction of the respective DCGL.

Confirmatory survey locations CU1-3C and CU1-3D were inaccessible at the time of the survey; therefore, no measurement or sample results are reported for either location in this report

8. COMPARISON OF RESULTS WITH GUIDELINES The confirmatory soil sample results were compared with the individual radionuclide DCGLs, and the SOF values were compared with the unity DCGL with the understanding that the SOF values may be underestimated since confirmatory analyses did not include quantification of the secondary ROCs (H-3, C-14, Ni-63, Pu-238, and Pu-239/240).

8.1 COMPARISON OF SURFACE ACTIVITY MEASUREMENTS The total surface activity values were directly compared with Appendix B of NUREG-1757, Consolidated Decommissioning Guidance (NRC 2006). The total and removable surface activity levels for all random and judgmental surface measurement locations were below the limiting DCGL for Co-60.

The average (mean) surface activities reported for the side-by-side results indicated a general agreement between the confirmatory and Enercon data. The surface activity results for side-by-side measurement locations are provided in Tables B-4 and B-5.

8.2 ANALYTICAL COMPARISON OF SAMPLES Confirmatory analyses of the four split samples identified measureable concentrations of Co-60, Cs-137, and Ag-108m in the Class 1 stockpile soil samples. The highest concentrations associated BMRC-Phase 2 Confirmatory Survey Report 13 5177-SR-02-1

with the split samples were 0.79 pCi/g of Ag-108m, 0.12 pCi/g of Co-60, and 0.07 pCi/g of Cs-137.

The individual results of the split sample analyses performed are provided in Table B-6.

Although ORAU had not received all Enercon final status survey data for review at the time of this report, the preliminary comparisons of the current data indicate that the ROC concentrations for confirmatory and FSS soil populations are similar.

9.

SUMMARY

At the NRCs request, ORAU conducted confirmatory surveys at the site of the former BMRC on the SUNY-UB South Campus in Buffalo, New York during the periods of January 26-29, February 3-6, and August 17-21, 2015. The survey activities included visual inspections, gamma radiation surface scans, gamma and beta radiation measurements, and soil sampling activities of six FSS units which were combined into two confirmatory survey units. Confirmatory activities also included the review and assessment of the licensees project documentation and methodologies.

The majority of gamma surface scans and total surface activity measurements were not distinguishable from background. Three judgmental and two split samples contained radionuclide concentrations above method MDCs, but all sample concentrations were below the respective DCGL and the unity rule for multiple radionuclides also met.

Based on the results of the confirmatory surveys, agreement between the mean surface activities of the FSS and confirmatory data, ORAU is of the opinion that Enercon has accurately and adequately demonstrated that SUs 1-6 of the BMRC site satisfy the approved license termination screening levels described in the licensees DP and FSSP.

BMRC-Phase 2 Confirmatory Survey Report 14 5177-SR-02-1

10. REFERENCES Enercon 2010. Characterization Plan, Buffalo Materials Research Center. Revision 0. Enercon. Murrayville, Pennsylvania. November 24.

Enercon 2011. Historical Site Assessment, Buffalo Materials Research Center. Revision 1. Enercon.

Murrayville, Pennsylvania. May 31.

Enercon 2012a. Decommissioning Plan, Buffalo Materials Research Center, Rev.1. Prepared for the State University of New York, University at Buffalo, Buffalo Material Research Center Office of Environment, Health, and Safety Services. Buffalo, New York. February 6.

Enercon 2012b. Final Status Survey Plan, Buffalo Materials Research Center, Rev.1. Prepared for the State University of New York, University at Buffalo, Buffalo Material Research Center Office of Environment, Health, and Safety Services. Buffalo, New York. September 20.

Enercon 2014. Letter to D. Vasbinder (UB). Re: Clarification Regarding Sampling and Surveying of Stockpiled Soil for BMRC Decommissioning. Enercon. Murrayville, Pennsylvania. November 19.

Enercon 2015. BMRC FSS Survey Package SU1 to SU4 Class I - Revision 3. Enercon. Murrayville, Pennsylvania. January.

NRC 2000. Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), NUREG-1575; Revision 1. U.S. Nuclear Regulatory Commission. Washington, DC. August.

NRC 2006. Consolidated Decommissioning Guidance: Decommissioning Processes for Material Licensee-Final Report, NUREG-1757; Revision 2. U.S. Nuclear Regulatory Commission. Washington, DC.

September 6.

ORAU 2013. Project-Specific Pan for Independent Confirmatory Survey Activities Associated with the Buffalo Materials Research Center at the State University of New York, University of Buffalo, Buffalo, New York.

ORAU. Oak Ridge, Tennessee. October 23.

ORAU 2014. Addendum to the Project-Specific Pan for Independent Confirmatory Survey Activities Associated with the Buffalo Materials Research Center at the State University of New York, University of Buffalo, Buffalo, New York. ORAU. Oak Ridge, Tennessee. December 18.

ORAU 2015a. ORAU Radiological and Environmental Survey Procedures Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. August 6.

ORAU 2015b. ORAU Environmental Services and Radiation Training Quality Program Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. August 7.

ORAU 2015c. ORAU Radiological and Environmental Analytical Laboratory Procedures Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. May 7.

ORAU 2015d. Email from E. Harpenau (ORAU) to T. Smith (NRC) and J. Clements (NRC),

RE: Revised Survey Packages for Units 1 through 4. ORAU. Oak Ridge, Tennessee. January 16.

BMRC-Phase 2 Confirmatory Survey Report 15 5177-SR-02-1

ORAU 2015e. ORAU Health and Safety Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. June.

ORAU/ORISE 2014. Radiation Protection Manual. Prepared by ORAU under the Oak Ridge Institute for Science and Education contract. Oak Ridge, Tennessee. October 14.

BMRC-Phase 2 Confirmatory Survey Report 16 5177-SR-02-1

APPENDIX A FIGURES BMRC-Phase 2 Confirmatory Survey Report 5177-SR-02-1

Figure A-1. Location of Buffalo Materials Research Center, Buffalo, New York BMRC-Phase 2 Confirmatory Survey Report A-1 5177-SR-02-1

Figure A-2. Confirmatory Units and Designated Classifications BMRC-Phase 2 Confirmatory Survey Report A-2 5177-SR-02-1

Figure A-3. Gamma Walkover Survey of Confirmatory Unit 1 BMRC-Phase 2 Confirmatory Survey Report A-3 5177-SR-02-1

Figure A-4. Gamma Walkover Survey of Confirmatory Unit 2 BMRC-Phase 2 Confirmatory Survey Report A-4 5177-SR-02-1

Figure A-5. Gamma Walkover Survey of Backfill Lift 1 and Adjacent Stockpile Footprint Area BMRC-Phase 2 Confirmatory Survey Report A-5 5177-SR-02-1

Figure A-6. Gamma Walkover Survey of Backfill Lift 2 BMRC-Phase 2 Confirmatory Survey Report A-6 5177-SR-02-1

Figure A-7. Gamma Walkover Survey of Stockpile Area BMRC-Phase 2 Confirmatory Survey Report A-7 5177-SR-02-1

Figure A-8. Planned Random Sample Locations in Confirmatory Units 1 and 2 BMRC-Phase 2 Confirmatory Survey Report A-8 5177-SR-02-1

Figure A-9. Actual Random Sample Locations in Confirmatory Units 1 and 2 BMRC-Phase 2 Confirmatory Survey Report A-9 5177-SR-02-1

Figure A-10. Judgmental Sample Locations in Confirmatory Units 1 and 2 BMRC-Phase 2 Confirmatory Survey Report A-10 5177-SR-02-1

Figure A-11. Side-by-Side Smear/Measurement Locations in Survey Unit 4 BMRC-Phase 2 Confirmatory Survey Report A-11 5177-SR-02-1

Figure A-12. Side-by-Side Measurement Locations in Survey Unit 6 BMRC-Phase 2 Confirmatory Survey Report A-12 5177-SR-02-1

APPENDIX B TABLES BMRC-Phase 2 Confirmatory Survey Report 5177-SR-02-1

Table B-1. Surface Activity Levels Gross Gamma Gross Beta Net Beta Surface Removable Activity Sample/Location ID Material Count Rate Count Rate Activity (dpm/100 cm2)

(cpm) (cpm) (dpm/100 cm2)ab Alpha Beta 5177R0022/CU1-1A Bedrock 8,494 337 550 -1 5 5177R0023/CU1-1B Bedrock 5,399 240 -420 -1 0 5177R0024/CU1-1C Bedrock 5,692 275 -70 1 5 5177R0025/CU1-1D Bedrock 4,221 247 -350 -1 -1 5177R0026/CU1-1E Bedrock 5,158 301 190 -1 2 5177R0027/CU1-2A Bedrock 4,709 327 450 -1 0 5177R0028/CU1-2B Bedrock 4,666 266 -160 -1 -2 5177R0029/CU1-2C Bedrock 4,079 308 260 3 -1 5177R0030/CU1-2D Bedrock 5,491 305 230 -1 5 5177R0031/CU1-2E Bedrock 5,154 349 670 -1 -4 5177R0032/CU1-3A Bedrock 4,849 245 -370 3 -2 5177R00033/CU1-3B Bedrock 4,392 260 -220 -1 2 Inaccessible/CU1-3C Bedrock NS NS Inaccessible/CU1-3D Bedrock NS NS 5177R0036/CU1-3E Bedrock 4,958 294 120 1 4 CU2-1A Soil 5,836 431 -590 CU2-1B Soil 6,027 419 -710 CU2-1C Soil 7,790 356 -1,300 CU2-1D Soil 8,841 409 -810 CU2-1E Soil 7,959 444 -460 CU2-2A Soil 7,823 404 -860 CU2-2B Soil 7,020 353 -1,400 CU2-2C Soil 6,167 307 -1,800 BMRC-Phase 2 Confirmatory Survey Report B-1 5177-SR-02-1

Table B-1. Surface Activity Levels Gross Gamma Gross Beta Net Beta Surface Removable Activity Sample/Location ID Material Count Rate Count Rate Activity (dpm/100 cm2)

(cpm) (cpm) (dpm/100 cm2)ab Alpha Beta CU2-2D Soil 8,273 506 160 CU2-2E Soil 7,604 424 -660 5177R0012/SU4-09 Bedrock 5,702 322 400 -1 -2 5177R0013/SU4-08 Bedrock 4,711 242 -400 -1 -1 5177R0014/SU4-14 Bedrock 4,567 268 -140 -1 1 5177R0015/SU4-12 Bedrock 4,838 254 -280 -1 -4 5177R0016/SU4-18 Bedrock 4,520 239 -430 -1 1 5177R0017/SU4-25 Bedrock 4,960 277 -50 -1 2 5177R0018/SU4-24 Bedrock 4,528 249 -330 3 -1 5177R0019/SU4-27 Bedrock 4,537 310 280 1 -1 5177R0020/SU4-28 Bedrock 4,897 323 410 -1 -1 5177R0021/SU4-19 Bedrock 4,294 252 -300 -1 -2 SU6-18 Soil 8,764 564 740 SU6-24 Soil 7,928 495 50 SU6-25 Soil 8,934 493 30 SU6-23 Soil 8,613 533 430 SU6-22 Soil 8,357 437 -530 SU6-17 Soil 8,898 637 1,500 SU6-12 Soil 9,250 636 1,500 SU6-15 Soil 6,927 495 50 SU6-16 Soil 8,156 572 820 SU6-11 Soil 8,530 664 1,700 SU6-10 Soil 8,252 592 1,000 BMRC-Phase 2 Confirmatory Survey Report B-2 5177-SR-02-1

Table B-1. Surface Activity Levels Gross Gamma Gross Beta Net Beta Surface Removable Activity Sample/Location ID Material Count Rate Count Rate Activity (dpm/100 cm2)

(cpm) (cpm) (dpm/100 cm2)ab Alpha Beta SU6-09 Soil 7,766 481 -90 SU6-04 Soil 7,455 559 690 SU6-05 Soil 9,594 657 1,700 SU6-06 Soil 8,902 617 1,300 SU6-01 Soil 4,546 374 -1,200 SU6-02 Soil 6,600 390 -1,000 Judgmental 1 (SU5)c Soil 95,525 624 1,300 -1 0 Judgmental 2 (SU4) Bedrock 3,761 246 -360 Judgmental 3 (SU4) Bedrock 3,431 266 -160 Judgmental 4 (SU4) Bedrock 3,326 243 -390 Judgmental 5 (SU4) Bedrock 3,518 257 -250 Judgmental 6 (SU4) Bedrock 3,444 224 -580 Judgmental 7 (SU4) Bedrock 3,911 250 -320 Judgmental 8 (SU6) Soil 9,256 636 1,500 Judgmental 9 (SU6) Soil 9,594 657 1,700 aSurface activity was calculated after subtracting the derived background value for bedrock or soil of 282 cpm and 490 cpm, respectively, from each gross measurement. Total efficiency of 0.10 for the Ludlum Model 44-142 scintillation detector was based on Co-60 (calibrated to Tc-99).

bValues have been rounded to two significant figures cJudgmental smear ID 5177R0034 BMRC-Phase 2 Confirmatory Survey Report B-3 5177-SR-02-1

Table B-2. Radionuclide Concentrations in Soilsa Ag-108m Am-241 Co-60 Cs-137 Eu-152 Eu-154 Sample IDb or SOF Statistic (pCi/g) Uncert.c (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert.

5177S0011 0.00 0.01 -0.02 0.03 0.01 0.02 0.05 0.01 -0.03 0.04 -0.01 0.02 -0.01 5177S0012 -0.01 0.01 -0.02 0.03 0.01 0.02 0.01 0.01 -0.01 0.04 0.01 0.02 -0.01 5177S0013 -0.01 0.01 -0.02 0.03 -0.02 0.02 -0.01 0.02 -0.01 0.05 0.02 0.02 -0.02 5177S0014 0.05 0.01 -0.06 0.03 0.01 0.02 0.02 0.01 0.00 0.04 0.00 0.02 -0.02 5177S0015 0.00 0.01 -0.01 0.03 0.00 0.02 0.01 0.01 0.00 0.04 0.00 0.02 0.00 5177S0016 -0.01 0.02 -0.03 0.03 0.01 0.02 0.01 0.01 0.02 0.04 0.02 0.02 -0.01 5177S0017 0.00 0.02 0.00 0.04 -0.01 0.02 -0.02 0.02 -0.02 0.04 0.00 0.03 -0.01 5177S0018 0.04 0.01 0.00 0.03 0.00 0.03 0.02 0.01 -0.04 0.05 0.01 0.02 0.00 5177S0019 0.00 0.01 -0.05 0.03 0.01 0.02 -0.01 0.02 -0.01 0.04 -0.02 0.02 -0.03 5177S0020 -0.01 0.01 -0.03 0.03 0.00 0.02 0.01 0.02 0.01 0.04 -0.01 0.02 -0.01 Mean -0.01 -0.02 0.01 0.03 -0.02 0.00 Median -0.01 -0.02 0.01 0.03 -0.02 0.00 St. Dev. 0.01 0.00 0.00 0.03 0.01 0.01 Minimum -0.01 -0.02 0.01 0.01 -0.03 -0.01 Maximum 0.00 -0.02 0.01 0.05 -0.01 0.01 aAnalysis for Sr-90 was not performed on Class 2 samples unless specifically requested by NRC.

bSamples collected from CU 2 c Uncert. = two sigma total propagated uncertainty is presented BMRC-Phase 2 Confirmatory Survey Report B-4 5177-SR-02-1

Table B-3. Radionuclide Concentrations for Judgmental Bedrock and Soil Samples Ag-108m Am-241 Co-60 Cs-137 Eu-152 Eu-154 Sr-90 Sample ID or SOF Statistic (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert.

5177M0001a 0.00 0.0 0.02 0.02 0.00 0.03 0.00 0.02 0.01 0.05 -0.01 0.02 -0.29 0.34 -0.16 5177M0002a -0.01 0.02 -0.04 0.03 -0.01 0.03 -0.03 0.03 0.00 0.06 -0.01 0.03 0.14 0.38 0.06 5177M0003a 0.02 0.02 0.02 0.03 0.00 0.03 0.12 0.03 0.05 0.06 0.01 0.02 -0.03 0.35 0.01 5177M0004a 0.00 0.02 -0.07 0.03 0.01 0.03 0.03 0.02 -0.04 0.05 -0.01 0.03 -0.23 0.31 -0.17 5177M0005a 0.02 0.01 -0.02 0.02 -0.01 0.01 -0.01 0.01 0.00 0.03 -0.01 0.01 -0.04 0.36 -0.04 5177M0006a 0.01 0.03 -0.04 0.04 -0.03 0.05 0.04 0.04 -0.01 0.08 -0.01 0.03 -0.32 0.35 -0.21 5177M0007b 0.01 0.0 0.02 0.04 -0.02 0.02 0.00 0.01 0.00 0.04 0.01 0.02 c 0.01 5177M0008b 0.02 0.01 -0.01 0.03 0.02 0.02 0.01 0.02 -0.01 0.04 -0.01 0.02 c 0.00 aSample collected from CU 1 bSample collected from SU 6 cSr-90 was not performed on Class 2 samples unless specifically requested by NRC.

Uncert. = two sigma total propagated uncertainty is presented BMRC-Phase 2 Confirmatory Survey Report B-5 5177-SR-02-1

Table B-4. Side-by-Side Measurements for Survey Unit 4 Enercon Gross Gamma Counts Gross Beta Counts Surface Activity Location (cpm) (cpm) (dpm/100 cm2)

Code ORAU Enercon ORAU Enercon ORAUa Enerconb SU4-09 5,702 5,269 322 83 400 0c SU4-08 4,711 5,018 242 88 -400 0 SU4-14 4,567 5,185 268 92 -140 0 SU4-12 4,838 5,072 254 81 -280 0 SU4-18 4,520 5,115 239 79 -430 0 SU4-25 4,960 4,896 277 79 -50 0 SU4-24 4,528 5,317 249 83 -330 0 SU4-27 4,537 5,056 310 94 280 0 SU4-28 4,897 5,150 323 91 410 0 SU4-19 4,294 4,968 252 82 -300 0 Mean: -84 0 aValues have been rounded to two significant figures.

bEnercon used incorrect surface efficiency for the Ludlum Model 43-93 detector.in calculation. Data are expected to be corrected in final report.

cEnercon converted negative activity values to zero (0).

BMRC-Phase 2 Confirmatory Survey Report B-6 5177-SR-02-1

Table B-5. Side-by-Side Measurements for Survey Unit 6 Gross Gamma Counts Gross Beta Counts Surface Activity Enercon Location (cpm) (cpm) (dpm/100 cm2)

Code ORAU Enercon ORAU Enercon ORAUa Enerconb SU6-18 8,764 7,414 564 78 740 538 SU6-24 7,928 7,147 495 55 50 0c SU6-25 8,934 8,200 493 87 30 793 SU6-23 8,613 7,414 533 78 430 538 SU6-22 8,357 7,292 437 84 -530 708 SU6-17 8,898 8,208 637 85 1,500 736 SU6-12 9,250 7,842 636 94 1,500 991 SU6-15 6,927 6,279 495 71 50 340 SU6-16 8,156 7,237 572 80 820 595 SU6-11 8,530 7,404 664 101 1,700 1,189 SU6-10 8,252 7,474 592 87 1,000 793 SU6-09 7,766 7,263 481 81 -90 623 SU6-04 7,455 7,355 559 73 690 396 SU6-05 9,594 9,265 657 90 1,700 878 SU6-06 8,902 7,668 617 110 1,300 1,444 SU6-01 4,546 8,096 374 61 -1,200 57 SU6-02 6,600 6,180 390 88 -1,000 821 Mean: 510 673 aValues have been rounded to two significant figures.

bEnercon used incorrect surface efficiency for the Ludlum Model 43-93 detector.in calculation. Data expected to be corrected in final report.

cEnercon converted negative activity values to zero (0).

BMRC-Phase 2 Confirmatory Survey Report B-7 5177-SR-02-1

Table B-6. Summary Statistics for Split Soil Samples Sample Ag-108m Am-241 Co-60 Cs-137 ID or ORAU Enercon ORAU Enercon ORAU Enercon ORAU Enercon Statistic (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert.

5177S0021a 2014 11 14 -0.01 0.01 -0.02 0.07 0.01 0.04 0.11 0.31 0.00 0.02 -0.01 0.39 0.05 0.02 0.03 0.05 S2-01-DUP 5177S0022b 2014 11 17 0.59 0.03 0.91 0.15 -0.06 0.03 0.24 0.37 0.08 0.02 0.10 0.07 0.07 0.02 0.15 0.09 S1-01-DUP 5177S0023b 2014 12 23 0.79 0.04 0.98 0.21 -0.01 0.04 -0.03 0.17 0.12 0.03 0.02 0.04 0.05 0.02 0.02 0.10 S1-01-DUP 5177S0024a 2014 12 29 0.00 0.02 0.02 0.03 -0.01 0.04 0.10 0.47 0.01 0.02 0.05 0.04 0.02 0.01 0.02 0.06 S2-01-DUP Mean 0.34 0.47 -0.02 0.10 0.05 0.04 0.05 0.05 Median 0.30 0.47 -0.01 0.11 0.05 0.04 0.05 0.03 St. Dev. 0.41 0.55 0.03 0.11 0.06 0.05 0.02 0.06 Minimum -0.01 -0.02 -0.06 -0.03 0.00 -0.01 0.02 0.02 Maximum 0.79 0.98 0.01 0.24 0.12 0.10 0.07 0.15 BMRC-Phase 2 Confirmatory Survey Report B-8 5177-SR-02-1

Table B-6. Summary Statistics for Split Soil Samples Eu-152 Eu-154 Sr-90 Sample ID or ORAU Enercon ORAU Enercon ORAU Enercon SOF Statistic (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert. (pCi/g) Uncert.

5177S0021a 2014 11 14 0.01 0.04 0.06 0.15 0.01 0.02 0.03 0.09 NAc -0.10 0.09 0.01 S2-01-DUP 5177S0022b 2014 11 17 0.00 0.05 -0.01 0.53 -0.01 0.02 0.11 0.11 0.02 0.33 0.25 0.13 0.08 S1-01-DUP 5177S0023b 2014 12 23 -0.03 0.06 -0.06 0.18 0.01 0.03 0.01 0.11 0.28 0.36 0.27 0.09 0.29 S1-01-DUP 5177S0024a 2014 12 29 -0.02 0.04 0.11 0.15 0.00 0.02 -0.02 0.13 NAc 0.06 0.12 0.00 S2-01-DUP Mean -0.01 0.02 0.00 0.03 0.15 0.12 Median -0.01 0.02 0.01 0.02 0.15 0.15 St. Dev. 0.02 0.08 0.01 0.05 0.18 0.17 Minimum -0.03 -0.06 -0.01 -0.02 0.02 -0.10 Maximum 0.01 0.11 0.01 0.11 0.28 0.27 aSample from Class 2 stockpile bSample from Class 1 stockpile cSr-90 was not performed on Class 2 samples unless specifically requested by NRC.

NA=not applicable Uncert. = two sigma uncertainty is presented BMRC-Phase 2 Confirmatory Survey Report B-9 5177-SR-02-1

APPENDIX C SURVEY AND ANALYTICAL PROCEDURES BMRC-Phase 2 Confirmatory Survey Report 5177-SR-02-1

C.1 PROJECT HEALTH AND SAFETY ORAU performed all survey activities in accordance with the ORAU/ORISE Radiation Protection Manual, the ORAU Health and Safety Manual, and the ORAU Radiological and Environmental Survey Procedures Manual (ORAU/ORISE 2014, ORAU 2015e, and ORAU 2015a). Should ORAU have identified a hazard not covered in the ORAU Survey Procedures Manual or the projects work-specific hazard checklist, work would not have been initiated or continued until it was addressed by an appropriate job hazard analysis and hazard controls.

The proposed survey and sampling procedures were evaluated to ensure that any hazards inherent to the procedures themselves were addressed in current job hazard analyses. Additionally, prior to performing work, a pre-job briefing and walkdown to identify hazards present were completed and discussed with field personnel.

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 ORAU documents:

  • ORAU Radiological and Environmental Survey Procedures Manual (ORAU 2015a)
  • ORAU Environmental Services and Radiation Training Quality Program Manual (ORAU 2015b)
  • ORAU Radiological and Environmental Analytical Laboratory Procedures Manual (ORAU 2015c)

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

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, NIST Radiochemistry BMRC-Phase 2 Confirmatory Survey Report C-1 5177-SR-02-1

Intercomparison Testing Program, and Intercomparison Testing Program Laboratory Quality Assurance Programs

  • Training and certification of all individuals performing procedures
  • Periodic internal and external audits Detectors used for assessing surface activity were calibrated in accordance with ISO-7503 1 recommendations. Total beta efficiencies (total) were determined for each instrument/detector combination and consisted of the product of the 2 instrument efficiency (i) and surface efficiency (s): total = i x s. ISO-7503 recommends an s of 0.25 for beta emitters with a maximum energy of less than 0.4 MeV and an s of 0.5 for maximum beta energies greater than 0.4 MeV.

Based on the data in Enercons final status survey plan to use the Co-60 screening level, an efficiency using Tc-99 as a surrogate was calculated. That total efficiency was 0.10 for the plastic scintillators used to quantify beta surface activity.

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

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

C.3.2 SURFACE ACTIVITY MEASUREMENTS Measurements of total beta and alpha surface activity levels were performed using hand-held scintillation detectors coupled to portable ratemeter-scalers. Count rates (cpm), which were integrated over one minute with the detector held in a static position, were converted to activity levels (dpm/100 cm2) by dividing the count rate by the total static efficiency (ixs) and correcting for the physical area of the detector, which is 100 cm2. ORAU calculated material-specific 1International Standard. ISO 7503-1, Evaluation of Surface Contamination - Part 1: Beta-emitters (maximum beta energy greater than 0.15 MeV) and alpha-emitters. August 1, 1988.

BMRC-Phase 2 Confirmatory Survey Report C-2 5177-SR-02-1

backgrounds for bedrock and surface soil by averaging the direct measurement data collected at each location evaluated during confirmatory survey activities. The average background value for bedrock (282 cpm) was used for the a priori MDC calculation for beta activity in the example below:

3 + 4.65

=

Where:

B = background tot = total efficiency G = geometry correction factor (1.0)

The a priori beta static MDC was approximately 811 dpm/100 cm2 using a total efficiency of 0.10 for Co-60.

C.3.3 SURFACE SAMPLING Soil samples were collected using a clean garden trowel and pick-axe, then transferred into a clean sample container by ORAU personnel. ORAU also collected their respective split samples from the remaining quantities Enercon had retained under custody. Enercon had already double-bagged the split samples. ORAU personnel labeled each in accordance with ORAU survey procedures completed the required custody documentation.

C.4 RADIOLOGICAL 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 or other appropriate container. The quantity placed in the beaker was chosen to reproduce the calibrated counting geometry. Net material weights were determined and the samples counted using intrinsic germanium detectors coupled to a pulse height analyzer system. Background and Compton stripping, peak search, peak identification, and concentration calculations were performed using the computer capabilities inherent in the analyzer system. All total absorption peaks (TAPs) associated with the ROCs were reviewed for consistency of activity. Minimum detectable concentrations for each radionuclide are available upon request.

Spectra were also reviewed for other identifiable TAPs.

BMRC-Phase 2 Confirmatory Survey Report C-3 5177-SR-02-1

Radionuclidea TAP (MeV) MDC (pCi/g)

Ag-108m 0.434 0.05 Am-241 0.060 0.11 Co-60 1.332 0.06 Cs-137 0.662 0.05 Eu-152 0.344 0.1 Eu-154b 0.123 0.05 aSpectra were also reviewed for other identifiable TAPs.

bThe Eu-154 723.30 keV peak not used due to interference from Ag-108m 722.9 keV peak.

C.4.2 Sr-90 ANALYSIS Bedrock and soil samples were dissolved by a combination of potassium hydrogen fluoride and pyrosulfate fusions. The fusion cake was dissolved and strontium was co-precipitated on lead sulfate.

The strontium was separated from residual calcium and lead by precipitating strontium sulfate from ethylenediaminetetraacetic acid at a pH of 4.0. Strontium was separated from barium by complexing the strontium in diethylenetriaminepentaacetic acid while precipitating barium as barium carbonate.

The strontium was ultimately converted to strontium carbonate and counted on a low-background gas proportional counter. The typical MDC of the procedure is 0.4 pCi/g for a one hour count time.

C.5 DETECTION LIMITS Detection limits, referred to as MDCs, were based on 95% confidence level via the NUREG-1507 method. 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.

BMRC-Phase 2 Confirmatory Survey Report C-4 5177-SR-02-1

APPENDIX D MAJOR INSTRUMENTATION BMRC-Phase 2 Confirmatory Survey Report 5177-SR-02-1

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

D.1 SCANNING AND MEASUREMENT INSTRUMENT/DETECTOR COMBINATIONS D.1.1 GAMMA Ludlum NaI Scintillation Detector Model 44-10, Crystal: 5.1 cm x 5.1 cm (Ludlum Measurements, Inc., Sweetwater, Texas) coupled to:

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

Trimble Data Logger (Trimble Navigation Limited, Sunnyvale, California)

D.1.2 BETA Ludlum Plastic Scintillation Detector Model 44-142, 100 cm2 physical area coupled to:

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

D.2 LABORATORY ANALYTICAL INSTRUMENTATION High-Purity, Extended Range Intrinsic Detector CANBERRA/Tennelec Model No: ERVDS30-25195 (Canberra, Meriden, Connecticut)

Used in conjunction with:

Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, Tennessee) and Multichannel Analyzer Canberras Gamma Software Dell Workstation (Canberra, Meriden, Connecticut)

High-Purity, Intrinsic Detector Model No. GMX-45200-5 CANBERRA Model No: GC4020 (Canberra, Meriden, Connecticut)

Used in conjunction with:

Lead Shield Model G-11 Lead Shield Model SPG-16-K8 (Nuclear Data)

Multichannel Analyzer Canberras Gamma Software BMRC-Phase 2 Confirmatory Survey Report D-1 5177-SR-02-1

Dell Workstation (Canberra, Meriden, Connecticut)

Low Background Gas Proportional Counter Model LB-5100-W (Tennelec/Canberra, Meriden, Connecticut)

Tri-Carb Liquid Scintillation Analyzer Model 3100 (Packard Instrument Co., Meriden, Connecticut)

BMRC-Phase 2 Confirmatory Survey Report D-2 5177-SR-02-1

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