ML14206A989

T. Smith Ltr Orau Independent Confirmatory Survey Summary and Results for the Buffalo Materials Research Center, New York

ML14206A989
Person / Time
Site:	05000157
Issue date:	07/24/2014
From:	Harpenau E Oak Ridge Associated Universities
To:	Tanya Smith NRC/FSME
References
5177-SR-01-1, RFTA 12-010
Download: ML14206A989 (56)

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INDEPENDENT CONFIRMATORYSURVEY

SUMMARY

AND RESULTS FOR THE BUFFALO MATERIALS RESEARCH CENTER BUFFALO, NEW YORK Evan M. Harpenau Prepared for U.S. Nuclear Regulatory Commission

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.

INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE BUFFALO MATERIALS RESEARCH CENTER BUFFALO, NEW YORK Prepared by E. M. Harpenau Independent Environmental Assessment and Verification Program Oak Ridge Institute for Science and Education Oak Ridge, Tennessee 37831-0017 Prepared for the U.S. Nuclear Regulatory Commission FINAL REPORT July 2014 Prepared by Oak Ridge Associated Universities 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.

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

1. INTRODUCTION....................................................................................................................................... 1

2. SITE DESCRIPTION................................................................................................................................. 3

3. OBJECTIVES................................................................................................................................................ 5

4. RADIONUCLIDES OF CONCERN...................................................................................................... 5 4.1 Release Criteria for Building Surfaces............................................................................................ 6 4.2 Release Criteria For Surface Soils................................................................................................... 6

5. PROCEDURES............................................................................................................................................ 7 5.1 Document Review............................................................................................................................. 7 5.2 Reference System.............................................................................................................................. 8 5.3 Surface Scans...................................................................................................................................... 8 5.4 Surface Activity Measurements....................................................................................................... 9 5.5 Soil Sampling...................................................................................................................................... 9

6. SAMPLE ANALYSIS AND DATA INTERPRETATION................................................................. 9

7. FINDINGS AND RESULTS................................................................................................................... 10 7.1 Document Review........................................................................................................................... 10 7.2 Surface Scans.................................................................................................................................... 11 7.3 Surface Activity Measurements..................................................................................................... 11 7.4 Radionuclide Concentrations In Soil Samples............................................................................ 12

8. COMPARISON OF RESULTS WITH GUIDELINES..................................................................... 12

9.

SUMMARY

.................................................................................................................................................. 13

10. REFERENCES......................................................................................................................................... 14 APPENDIX A. FIGURES APPENDIX B. SCAN DATA APPENDIX C. TABLES APPENDIX D. MAJOR INSTRUMENTATION APPENDIX E. SURVEY AND ANALYTICAL PROCEDURES BMRC Confirmatory Survey Report ii 5177-SR-01-0

TABLES Table 2.1. Areas Investigated During the Confirmatory Survey................................................................. 4 Table 4.1. BMRC Radionuclides of Concerna............................................................................................... 5 Table 4.2. DCGLs for Primary Radionuclides of Concern in Soila............................................................ 6 Table 7.1. Summary of Concentrations in the Tank Farm Excavation................................................... 12 Table C-1. Judgmental Measurement Locations for the Neutron Deck in the Reactor Containment Building........................................................................................................................................ C-1 Table C-2. Judgmental Measurement Locations for the Gamma Deck in the Reactor Containment Building........................................................................................................................................ C-2 Table C-3. Judgmental Measurement Locations Associated With the Hot Cell................................. C-3 Table C-4. Judgmental Measurement Locations for the Gamma Deck in the Administration Building........................................................................................................................................ C-4 Table C-5. Judgmental Measurement Locations for the Control Deck in the Reactor Containment Building........................................................................................................................................ C-5 Table C-6. Radionuclide Concentrations in Soil (pCi/g)........................................................................ C-6 FIGURES Fig. 2.2. Interior Cutaway View of the Reactor and Containment............................................................. 4 Fig. A-1. Location of Buffalo Materials Research Center, Buffalo, New York................................... A-1 Fig. A-2. BMRC, Control Deck - Scan Coverage.................................................................................... A-2 Fig. A-3. BMRC, Gamma Deck - Scan Coverage................................................................................... A-3 Fig. A-4. BMRC, Neutron Deck - Scan Coverage.................................................................................. A-4 Fig. A-5. BMRC, Control Deck - Direct Measurement Locations....................................................... A-5 Fig. A-6. BMRC, Gamma Deck - Direct Measurement Locations...................................................... A-6 Fig. A-7. BMRC, Neutron Deck - Direct Measurement Locations..................................................... A-7 Fig. A-8. BMRC, Exterior Soil Surfaces - Soil Sample Locations......................................................... A-8 BMRC Confirmatory Survey Report iii 5177-SR-01-0

ACRONYMS BMRC Buffalo Materials Research Center BSFR bulk survey for release CFR Code of Federal Regulations cpm counts per minute DCGLW derived concentration guideline level DM direct measurement DP decommissioning plan dpm/100 cm2 disintegrations per minute per one hundred square centimeters FSS final status survey FSSP final status survey plan IEAV Independent Environmental Assessment and Verification INEEL Idaho National Engineering and Environmental Laboratory ISM Integrated Safety Management MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDC minimum detectable concentration mrem/yr millirem per year NaI sodium iodide NIST National Institute of Standards and Technology NRC U.S. Nuclear Regulatory Commission ORAU Oak Ridge Associated Universities ORISE Oak Ridge Institute for Science and Education pCi/g picocuries per gram ROC radionuclide of concern SOF sum of fractions SSC structures, systems, and components SUNY State University of New York TAP total absorption peak UB University at Buffalo BMRC Confirmatory Survey Report iv 5177-SR-01-0

INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE BUFFALO MATERIALS RESEARCH CENTER BUFFALO, NEW YORK

1. INTRODUCTION The Buffalo Materials Research Center (BMRC) 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 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).

The reactor at the BMRC was operated under Atomic Energy Commissionpredecessor to the U.S.

Nuclear Regulatory Commission (NRC)License Number R-77 from 1961 to 1963 as a materials-testing reactor. In 1964, the reactor was shut down and the core and control systems were modified so that the reactor could operate with Pulse Training Assembled Reactor fuel at power levels up to 2 megawatts.

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. Decommissioning activities will result in the complete removal of BMRC Confirmatory Survey Report 1

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the BMRC structures from the site, allowing an unrestricted release of the BMRC site by the NRC and termination of the NRC license per Title 10, Code of Federal Regulations, Part 20, Section 1401 (10 CFR 20).

All structures of the facility are planned to be removed in their entirety. Most of the reactor components and systems are either activated or contaminated and will be segregated from non-radiological components and surfaces so that they can be disposed of as low-level radioactive waste. The BMRC is constructed primarily of concrete and a vast majority of the waste generated during decommissioning will be concrete rubble. Based on characterization results, most of the waste is planned to be sent for concrete recycling or sent to an industrial landfill. Some concrete, primarily the floor slabs of the subbasement and the Neutron Deck, will be disposed of as bulk survey for release (BSFR) waste due to the potential for volumetric contamination. The rubble waste will be sent to Studsvik, which will assess the waste to determine the disposal pathway. Materials deemed as meeting the BSFR requirements will be disposed of in an industrial landfill. Materials that do not meet the BSFR requirements will be sent to a low-level radioactive waste site (Studsvik 2013).

Targeted decontamination may be required for some reactor components, the Bioshield, and the hot cell concrete structures that were outside of the main neutron activation zone. However, the vast majority of the concrete and metal from these structures are expected to meet the requirements of BSFR waste. After building structural surfaces have been decontaminated and all radioactive wastes are disposed of, the interior structural surfaces will be released using the guidelines provided in IE Circular 81-07 (Enercon 2012a).

Upon completion of the removal of all the structures, a final status survey (FSS) of the footprint is required to demonstrate compliance with default screening values for soil (NRC 1998a and 10 CFR 20) and for bedrock/building surfaces (NRC 2006) in support of the unrestricted release of the site and license termination. The footprint requiring FSS will be a 15-20 foot excavation with a bedrock floor and soil sidewalls. The BMRC FSS will include beta walkover scans and direct measurements of the bedrock surfaces and gamma walkover and soil sampling over soil surfaces for radionuclides of concern (ROCs) (Enercon 2012b).

At the NRCs request, the Independent Environmental Assessment and Verification (IEAV)

Program of Oak Ridge Associated Universities (ORAU) conducted confirmatory survey activities at BMRC Confirmatory Survey Report 2

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the BMRC to verify the licensees surveys will permit demolition of the buildings. The licensee had already completed their free release surveys of the building at the time of the confirmatory survey.

The confirmatory activities were performed under the Oak Ridge Institute for Science and Education (ORISE) contract, which is managed and operated by ORAU for the U.S. Department of Energy.

2. SITE DESCRIPTION Located on the southern edge of the South Campus of the UB off of Rotary Drive in Buffalo, Erie County, New York (Fig. A-1), the site is approximately 20 miles south of the Canadian border at Niagara Falls, New York, 90 miles west of Rochester, and 80 miles from the southern border of New York and Pennsylvania. UB is also approximately five miles east of the Niagara River, which borders Canada.

The BMRC Facility consists of the reactor, the Containment Building which encloses 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 contains offices, classrooms, and laboratories. The Containment Building is constructed of reinforced concrete in a right cylindrical shape and is approximately 75 feet in diameter and 52 feet high with two-foot thick walls. The walls and the foundation were constructed on bedrock. Floor plans for each of the three decks are shown in Figures A-2 through A-4 in Appendix A.

The Containment Building has three levels: the Control Deck, the Gamma Deck, and the Neutron Deck (Fig. 2.2). The areas investigated by ORAU during the confirmatory survey are outlined in Table 2.1.

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Table 2.1. Areas Investigated During the Confirmatory Survey Level Building Includes Licensee Classification Neutron ADM N02, N04, N06, SN06, 10k Vertical tank room 3

N03 2

RBC N01, High radiation area 2

Gamma ADM Rooms 115, 115A, 118, 114 3

RBC All rooms including hot cell 2

Control RBC All except Room 202 2

Room 202 1

ADM = Administration RBC = Reactor Building Containment Fig. 2.2. Interior Cutaway View of the Reactor and Containment Control Deck Gamma Deck Neutron Deck BMRC Confirmatory Survey Report 4

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3. OBJECTIVES The objectives of the confirmatory surveys 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 building release survey results.

4. RADIONUCLIDES OF CONCERN The primary ROCs for the BMRC are beta-gamma emittersfission and activation products resulting from reactor operation. Table 4.1 provides a comprehensive list of the ROCs for the BMRC and the areas of concern/matrix (Enercon 2012b).

Table 4.1. BMRC Radionuclides of Concerna Radionuclide Half-life (years)

Emission Area(s) of Concern Ag-108m 438 Soil; tank water; SSCsb Am-241 432 Tank sediment C-14 5,730 Laboratory areas Co-60 5.27 Soil; SSCs; Bioshield Cs-137 30.1

,c 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 2012a.

bSSCs = structures, systems, and components c emission from Ba-137m progeny BMRC Confirmatory Survey Report 5

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4.1 RELEASE CRITERIA FOR BUILDING SURFACES There are no dose-based or administrative release limits for building surfaces, instead the licensee has committed to remediate areas that are considered to contain detectable contamination, per IE Circular 81-07 guidance. As described in IE Circular 81-07, detectable levels of contamination are defined as total beta-gamma surface activity levels above 5,000 dpm/100cm2 and removable above 1,000 dpm/100cm2, based on the detection sensitivity of common survey instruments at the time of publication (NRC 1981). The licensee is assuming all surface contamination is due to cobalt-60 (Co-60), which is a conservative assumption because the radiation detector total efficiency is lower for Co-60 than the other ROCs, excluding the hard-to-detects (H-3, C-14, and Ni-63). After acceptable surface contamination levels have been attained, the remaining building structures will be deconstructed and the deconstruction materials (mainly concrete) will be sent to Studsvik, the BSFR waste processor for determination of the proper disposal pathways.

4.2 RELEASE CRITERIA FOR SURFACE SOILS Table 4.2 lists the screening values applicable to surface soils, and the selected derived concentration guideline levels (DCGLs) associated with the ROC for soil surfaces.

Table 4.2. DCGLs for Primary Radionuclides of Concern in Soila Radionuclide NRC Screening Value for Surface Soils (pCi/g)

Selected DCGL Value (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. Because each of the individual DCGL represents 25 mrem/yr, the sum-of-fractions (SOF) approach is used to demonstrate compliance with the dose limit. SOF calculations are performed as follows:

SOFTOTAL = SOFj =

=0

Cj DCGL,j

=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.

Soil concentration compliance will be demonstrated after building demolition and surveys of the footprint are performed.

5. PROCEDURES During the period of October 29 through October 31, 2013 ORAU performed a radiological survey of the BMRC. The survey was conducted in accordance with a confirmatory project-specific plan (PSP) and the ORAU/ORISE Survey Procedures and ORAU Quality Program Manual (ORAU/ORISE 2013a, 2014a, and ORAU 2014a). This report summarizes the procedures and results of the survey.

5.1 DOCUMENT REVIEW Prior to on-site activities, ORAU reviewed the licensees characterization plan, historical site assessment, decommissioning plan, and final status survey plan (Enercon 2010, 2011, 2012a, and 2012b). All documents and data were reviewed for adequacy and appropriateness while taking into account the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) guidance (NRC 2000). While onsite, ORAU staff reviewed the licensees final survey data packages for areas inside the reactor containment building.

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ORAU also reviewed additional data package submittals and licensee responses to comments (UB 2014 and Enercon 2014a, 2014b).

5.2 REFERENCE SYSTEM Interior structural direct measurements (DMs) and scans and exterior soil sample locations were referenced to prominent site features and documented on site drawings provided by the licensee.

5.3 SURFACE SCANS Surface scans for alpha-plus-beta and gamma radiation were performed using gas proportional and sodium iodide (NaI) scintillation detectors, respectively. Scan density was dependent on classification: Class 1 areas received high-density scans, Class 2 areas received medium-density scans, and Class 3 received low-density scans. During the surface scans, particular attention was given to cracks, joints, and horizontal surfaces where material may have accumulated. Detectors were coupled to ratemeter-scalers with audible indicators and were also coupled to dataloggers to electronically record all scan data at one-second intervals. Locations of elevated direct radiation that were audibly distinguishable from background levels, suggesting the presence of residual contamination, were marked for further investigation. Figures A-2 through A-4 indicate scan coverage of surface scans performed on building surfaces.

Portions of the bioshield on the neutron deck were not removed due to structural concerns; all areas of the bioshield that will be disposed of as radiological waste were marked with orange spray paint.

As a result, significant gamma shine was observed with all radiation detection equipment. Because of the gamma shine, it was not useful to scan the reactor pit area. In lieu of scanning the reactor pit, the ORAU survey team evaluated the boundary of the contaminated bioshield by placing lead bricks around the detector and scanning on either side of the boundary.

As shown in Table C-1, Neutron Deck, initial building tour, NRC identified an area containing concrete shield blocks that had not been surveyed by the licensee. The licensee remediated this area and performed surveys of the area and ORAU performed confirmatory surveys.

Gamma walkover scans were performed by NRC staff on exterior soil surfaces surrounding the BMRC and in the tank farm excavation. Scans were performed using NaI scintillation detectors BMRC Confirmatory Survey Report 8

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coupled to ratemeter-scalers with audible indicators. These surveys were performed to determine if radiological postings were required in the excavation area and to confirm that the other soils had not been impacted.

5.4 SURFACE ACTIVITY MEASUREMENTS Direct measurements were performed using gas proportional detectors coupled to portable ratemeter-scalers. Direct measurement locations are identified in Figures A-5 through A-7.

Construction material-specific background measurements were collected from the neutron deck in the reactor containment building on poured concrete and concrete block for correcting gross activity measurements performed on structural surfaces. The PSP specified that confirmatory direct measurement locations would be selected based on a random/systematic method. However, a deviation to the plan was necessary as the survey progressed such that all measurement locations were selected judgmentally based on surface scan results. A second deviation to the PSP was implemented, after discussion with NRC staff, to focus resources on conducting additional surface scans rather than performing side-by-side measurements and measurements for hard-to-detects.

These additional scans resulted in most of the reactor building surfaces receiving confirmatory investigations.

5.5 SOIL SAMPLING Ten surface soil samples were collected from the exterior land areas by NRC staff. Three samples were collected from the tank farm excavation (burial pit) on the south side of the containment building to verify radiological postings were adequate. The licensee planned to perform additional remediation of tank farm excavation area. The remaining samples were collected from non-impacted areas around the building. Selected sample locations were based on elevated direct gamma radiation levels identified during scans. Figure A-8 shows the approximate confirmatory sample locations.

6. SAMPLE ANALYSIS AND DATA INTERPRETATION Samples were returned to the ORAU/ORISE Radiological and Environmental Analytical Laboratory in Oak Ridge, Tennessee for analysis and interpretation. Sample analyses were performed in accordance with the ORAU/ORISE Laboratory Procedures Manual BMRC Confirmatory Survey Report 9

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(ORAU/ORISE 2014b). Soil samples were analyzed by gamma spectroscopy for gamma-emitting ROCs with results reported in units of picocuries per gram (pCi/g). Samples from the burial pit excavation (from an area with known Sr-90 contamination) and three additional samples from the soil surrounding the BMRC (not suspected to contain Sr-90) were analyzed for Sr-90. 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 for each of the confirmatory activities are discussed below.

7.1 DOCUMENT REVIEW The ORAU reviews of project documentation led to several pre-survey concerns that were addressed to the NRC in an e-mail dated October 23, 2013 (ORAU/ORISE 2013b). These concerns were addressed in a conference call with the NRC and the licensee on October 24, 2013. After reviewing the survey data packages, it was determined that the licensee had applied an inappropriate surface efficiency to the total efficiency calculation. When assuming all of the contamination is due to Co-60, the surface efficiency must reflect the contaminant. A surface efficiency of 0.25, instead of 0.5, should have been used per ISO 7503-1 (ISO 1998). The licensee submitted a revised data table (no date) and revised data packagesdated June, 5 and June 25, 2014where all surface activity data had been recalculated using a surface efficiency of 0.25. These data in the revised data packages showed that identified residual contamination had either been remediated or would be addressed during building demolition, and that all other surfaces were less than the building surface clearance limits. However, review of the revised data packages generated additional comments regarding inconsistencies, instrumentation and documentation data quality, and documentation of changes/revisions to data. A letter detailing the specific comments was submitted to the NRC as a separate document (ORAU 2014b). Other observations made during the confirmatory survey site visit are described below.

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7.2 SURFACE SCANS Surface scans of the selected areas within the BMRC identified several areas of elevated direct gamma and beta radiation. Specifically, scans identified elevated direct radiation at the former drain line on the neutron deck; on the hot cell and hot cell door on the gamma deck; and in Rooms 101A, 107, 109, and 202 of the RBC. Elevated direct radiation was also identified in Rm 115A of the ADM. In addition, several concrete blocks that were being stored outside of the 10k vertical tank room had an instrument response in excess of the action level. As a result, the licensee scanned each block and disposed of a total of nine blocks as radiological waste and was required to upgrade the classification of multiple survey units to Class 1 and perform additional surveys. After scan surveys had identified and bounded the elevated radiation, total surface activity measurements were performed to quantify the levels of contamination. Results for the surface scans performed on structural surfaces are presented in Appendix B. Appendix C shows the areas where scans identified and direct measurements confirmed surface activity levels in excess of the release guideline. Also presented in Appendix C, are the associated total and removable surface activity measurement data.

7.3 SURFACE ACTIVITY MEASUREMENTS Total surface activity measurements were collected from areas where direct radiation levels exceeded the scanning action level, as identified during scanning. The reported surface activities represent net levels that have had background contributions subtracted. Smear samples, to determine removable gross beta activity levels, were also collected from the surface measurement locations. Appendix C provides a summary of the radiological measurement data for each judgmental location. An area in Room 115A had direct radiation levels near the action level. Even though the area was below the criteria of IE Circular 81-07, the licensee remediated the contaminated area as directed by NRC.

The four locations of elevated activity identified in Room 202 were initially assumed by the licensee to be due to Sr-90 only; therefore, a Sr-90 total efficiency was used for the data conversion for measurements in this room. As a result of that assumption, two of the four locations were remediated by the licensee. The two remaining locations (with an approximate size of 300 cm2 in Room 202) were not considered detectable. However, after the confirmatory survey, the licensee decided they could not support the decision that all contamination was due to Sr-90; thus, a Co-60 efficiency was applied to all data. Applying the Co-60 instrument efficiency resulted in at least one of BMRC Confirmatory Survey Report 11 5177-SR-01-0

the two remaining locations exceeding the release criterion. Remediation of this location was required, as presented in Table C-5.

ORAU conducted surveys in the reactor building and NRC performed surveys in the remainder of the facility offices and support facilities. NRC identified suspected elevated measurements in rooms 115/115A and had ORAU perform detailed confirmatory surveys of this area.

7.4 RADIONUCLIDE CONCENTRATIONS IN SOIL SAMPLES Individual sample results for the gamma-emitting fission/activation products that the licensee identified as site-related contaminants are presented in Table C-6. Samples (5177S0001 through 5177S0003) collected from the tank farm excavation (burial pit) did contain elevated concentrations of Ag-108m, Cs-137, Co-60, and Sr-90 as summarized in Table 7.1 below.

Table 7.1. Summary of Concentrations in the Tank Farm Excavation Radionuclide Selected DCGL Value (pCi/g)

Concentration Range (pCi/g)

Ag-108m 8.2 7.7 to 13.5 Co-60 3.8 1.8 to 3.1 Cs-137 11 0.4 to 0.6 Sr-90 1.7 3.8 to 7.9 Samples taken from soil surrounding the BMRC (samples 5177S0004 to 5177S0010), excluding the tank farm excavation (burial pit), did not indicate the presence of any elevated concentrations of site-related contaminants.

8. COMPARISON OF RESULTS WITH GUIDELINES The total surface activity values were directly compared with the detection sensitivity outlined in IE Circular 81-07. All areas identified as having detectable surface contamination, were remediated while the ORAU survey team was onsite. The final surface activity measurements, performed after remediation efforts by the licensee, were free of detectable radioactivity in accordance with IE Circular 81-07.

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9.

SUMMARY

At NRCs request, ORAU conducted confirmatory radiological surveys of the BMRC during the period of October 29 through October 31, 2013. The survey activities included visual inspections and measurement and sampling activities. Activities also included the review and assessment of the licensees project documentation and methodologies.

ORAU noted two primary areas of concern once confirmatory survey activities were under way. The most significant finding was that all measurement locations above the levels specified in IE Circular 81-07, and thus remediated, were in a Class 2 or even Class 3 survey area. The MARSSIM provides very specific guidance as to classification of survey areas. Class 2 and 3 areas are not expected to have contamination above the release limit. Therefore, ORAU was of the opinion that the licensee mis-classified several survey areas, resulting in inadequate scan coverage. This was discussed with the NRC which then led to the licensee re-classifying and re-surveying several survey areas. Any areas of contamination that were determined to be detectable, per IE Circular 81-07 standards, during the re-survey were marked for remediation/removal during building demolition.

The second item of concern, as discussed in Section 7.1, was the inappropriate use of 0.5 as a surface efficiency for Co-60instead of the correct 0.25 value. Using a 0.5 surface efficiency would result in an action level that is too high. Also, using a 0.5 surface efficiency would decrease the static minimum detectable concentration (MDC) by a factor of two. Using the incorrect surface efficiency coupled with the inadequate scan coverage would cause the licensee to overlook several areas above the release criteria of IE Circular 81-07, as indicated by the results of this confirmatory survey. The licensee has since revised all surface activity calculations using a 0.25 surface efficiency. Additional potential issues with the instrumentation used for the release surveys were identified in the June 5 and June 25, 2014 data package submittals.

Based on the high-density confirmatory survey surface scans, the licensees remediation of residual contamination identified during the confirmatory surveys, and the licensees additional surveys and supporting documentation, all reported total and removable surface activity levels are below the criteria established in IE Circular 81-07. Exceptions are documented structural components or inaccessible components that will be removed during demolition.

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10. REFERENCES 10 CFR 20. Radiological Criteria for License Termination. Code of Federal Regulations, Chapter 10, Part 20 (10 CFR 20), Subpart E.

AEC 1974. Termination of Operating Licenses for Nuclear Reactors. Regulatory Guide 1.86. U.S. Atomic Energy Commission. Washington, DC. June.

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. Revision 1. Enercon.

Murrayville, Pennsylvania. May 31.

Enercon 2012b. Final Status Survey Plan, Buffalo Materials Research Center. Revision 1. Enercon.

Murrayville, Pennsylvania. September 20.

Enercon 2014a. Letter to D. Vasbinder (UB). Re: Response to June 2, 2014 Comments. Enercon.

Murrayville, Pennsylvania. June 5.

Enercon 2014b. Letter to D. Vasbinder (UB). Re: BMRC Building Release. Enercon. Murrayville, Pennsylvania. June 25.

Gilbert 1987. Statistical Methods for Environmental Pollution Monitoring. Van Nostrand Reinhold. New York, NY. Copyright 1987.

ISO 1998. Evaluation of surface contamination - Part 1: Beta-emitters (maximum beta energy greater than 0, 15 MeV) and alpha-emitters. International Organization for Standardization. ISO-7503-1; first edition. 1998 NRC 1981. Control of Radioactively Contaminated Material. IE Circular No. 81-07. U.S. Nuclear Regulatory Commission. Washington, DC. May 14.

NRC 1998a. Supplemental Information on the Implementation of the Final Rule on Radiological Criteria for License Termination. Federal Register, Volume 63, Number 222. U.S. Nuclear Regulatory Commission.

Washington, DC. November 18.

NRC 1998b. Minimum Detectable Concentrations With Typical Radiation Survey Instruments for Various Contaminants and Field Conditions. NUREG-1507. Washington, DC. June.

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.

BMRC Confirmatory Survey Report 14 5177-SR-01-0

ORAU 2012. Health and Safety Manual. Revision 16. Oak Ridge Associated Universities. Oak Ridge, Tennessee. May 12.

ORAU 2014a. Quality Program Manual for the Independent Environmental Assessment and Verification Program. Revision 29. Oak Ridge Associated Universities. Oak Ridge, Tennessee. June 4.

ORAU 2014b. Comments on the Building Demolition Survey Packages, Buffalo Materials Research Center at the State University of New York, University of Buffalo. DCN 5177-DR-01-0. Oak Ridge Associated Universities. Oak Ridge, Tennessee. July 16.

ORAU/ORISE 2011. Radiation Protection Manual. Oak Ridge Institute for Science and Education, managed and operated by Oak Ridge Associated Universities. Oak Ridge, Tennessee. December 3.

ORAU/ORISE 2013a. Project-Specific Plan for Independent Confirmatory Survey Activities Associated with the Buffalo Materials Research Center at the State University of New York, University of Buffalo, New York; DCN 5177-PL-01-0 (Docket No. 50-157; RFTA No.12-010). Oak Ridge Institute for Science and Education, managed and operated by Oak Ridge Associated Universities. Oak Ridge, Tennessee.

October 23.

ORAU/ORISE 2013b. E-mail from W. Adams (ORAU/ORISE) to T. Smith (NRC) RE: ORAU BMRC Survey Questions. Oak Ridge Institute for Science and Education, managed and operated by Oak Ridge Associated Universities. Oak Ridge, Tennessee. October 23.

ORAU/ORISE 2014a. Survey Procedures Manual for the Independent Environmental Assessment and Verification Program. Revision 23. Oak Ridge Institute for Science and Education, managed and operated by Oak Ridge Associated Universities. Oak Ridge, Tennessee. April 10.

ORAU/ORISE 2014b. Laboratory Procedures Manual for the Independent Environmental Assessment and Verification Program. Revision 54. Oak Ridge Institute for Science and Education, managed and operated by Oak Ridge Associated Universities. Oak Ridge, Tennessee. April 1.

Studsvik 2013. http://www.studsvik.com/en/Business-Areas/Waste-Treatment/Processing-of-Radioactive-Waste/Bulk-Survey-for-Release-BSFR/

University of Buffalo (UB) 2014. Email correspondence from D. Vasbinder (UB) to T. Smith (NRC),

Subject:

Response on Information for ORISE Confirmatory Survey Questions. March 11.

BMRC Confirmatory Survey Report 15 5177-SR-01-0

APPENDIX A FIGURES BMRC Radiological Survey Report 5177-SR-01-0

Fig. A-1. Location of Buffalo Materials Research Center, Buffalo, New York BMRC Radiological Survey Report A-1 5177-SR-01-0

Fig. A-2. BMRC, Control Deck - Scan Coverage BMRC Radiological Survey Report A-2 5177-SR-01-0

Fig. A-3. BMRC, Gamma Deck - Scan Coverage BMRC Radiological Survey Report A-3 5177-SR-01-0

Fig. A-4. BMRC, Neutron Deck - Scan Coverage BMRC Radiological Survey Report A-4 5177-SR-01-0

Fig. A-5. BMRC, Control Deck - Direct Measurement Locations BMRC Radiological Survey Report A-5 5177-SR-01-0

Fig. A-6. BMRC, Gamma Deck - Direct Measurement Locations BMRC Radiological Survey Report A-6 5177-SR-01-0

Fig. A-7. BMRC, Neutron Deck - Direct Measurement Locations BMRC Radiological Survey Report A-7 5177-SR-01-0

Fig. A-8. BMRC, Exterior Soil Surfaces - Soil Sample Locations BMRC Radiological Survey Report A-8 5177-SR-01-0

APPENDIX B SCAN DATA BMRC Radiological Survey Report 5177-SR-01-0

Surface Min Max Mean Median SD Lower Walls 201 682 338 320 75 Floor 902 3,429 1,195 1,113 310 Surface Min Max Mean Median SD Lower Walls 6,004 13,532 9,484 9,292 1,346 Floor 594a 40,158 11,066 10,341 3,473 Summary Statistics for Alpha-Plus-Beta Scans Summary Statistics for Gamma Scans Surface Scans for the Neutron Deck in the Reactor Containment Building aFrom the Q-plot there are 4 outlier present Due to shine from bioshield and hotspot identified under former drain line Due to shine from bioshield and hotspot identified under former drain line BMRC Confirmatory Survey Report B-1 5177-SR-01-1

Surface Min Max Mean Median SD Lower Walls 150 913 296 289 63 Floor 545 6,603 941 810 581 Surface Min Max Mean Median SD Lower Walls 4,659 30,976 7,859 7,526 1,848 Floor 3,772 21,371 8,133 8,009 1,724 Surface Scans for the Neutron Deck in the Adminstration Building Summary Statistics for Alpha-Plus-Beta Scans Summary Statistics for Gamma Scans From an area in the soutwest corner of N02 that will be disposed of as rad waste From an area in the southwest corner of N02 that will be disposed of as rad waste BMRC Confirmatory Survey Report B-2 5177-SR-01-1

Surface Min Max Mean Median SD Lower Walls 225 11,397 444 350 647 Floor 686 2,099 822 796 132 Surface Min Max Mean Median SD Lower Walls 4,920 10,489 7,210 7,185 828 Floor 5,424 10,664 7,485 7,521 748 Surface Scans for the Gamma Deck in the Reactor Containment Building Summary Statistics for Alpha-Plus-Beta Scans Summary Statistics for Gamma Scans Contamination identified on an electrical outlet in Rm 107 Note: Gamma scans were performed after remediation took place.

Elevated activity identified in Rm 101A; confirmed above limit with hand-held gas porportional BMRC Confirmatory Survey Report B-3 5177-SR-01-1

Surface Min Max Mean Median SD Hot Cell 149 6,129 866 543 936 Surface Min Max Mean Median SD Hot Cell 3,335 13,821 5,472 4,980 1,946 Surface Scans for the Hot Cell on the Gamma Deck Summary Statistics for Alpha-Plus-Beta Scans Summary Statistics for Gamma Scans BMRC Confirmatory Survey Report B-4 5177-SR-01-1

Surface Min Max Mean Median SD Lower Walls 194 1,012 357 345 65 Floor 1,145 2,946 1,623 1,567 244 Surface Min Max Mean Median SD Lower Walls 3,549 14,226 7,525 7,279 1,646 Floor 5,316 10,302 7,480 7,509 949 Surface Scans for the Gamma Deck in the Adminstration Building Summary Statistics for Alpha-Plus-Beta Scans Summary Statistics for Gamma Scans BMRC Confirmatory Survey Report B-5 5177-SR-01-1

Surface Min Max Mean Median SD Lower Walls 182 5,625 534 377 518 Floor 336 1,843 726 485 406 Surface Min Max Mean Median SD Lower Walls 5,314 9,874 7,827 7,785 790 Floor 6,578 9,580 7,942 7,942 616 Surface Scans for Rm 202 in the Reactor Containment Building Summary Statistics for Alpha-Plus-Beta Scans Summary Statistics for Gamma Scans From an area in the soutwest corner of N02 that will be disposed of as rad waste BMRC Confirmatory Survey Report B-6 5177-SR-01-1

Surface Min Max Mean Median SD Lower Walls 260 772 433 397 106 Floor 772 1,736 937 894 155 Surface Min Max Mean Median SD Lower Walls 5,186 11,927 7,276 7,081 1,103 Floor 5,220 10,975 7,746 7,726 855 Surface Scans for the Control Deck in the Reactor Containment Building Summary Statistics for Alpha-Plus-Beta Scans Summary Statistics for Gamma Scans From an area in the soutwest corner of N02 that will be BMRC Confirmatory Survey Report B-7 5177-SR-01-1

APPENDIX C TABLES BMRC Radiological Survey Report 5177-SR-01-0

Table C-1. Judgmental Measurement Locations for the Neutron Deck in the Reactor Containment Building Building Deck Area Reactor Building Containment Neutron Former drain line Direct Measurement Number Pre-Remediation Gross Beta Surface Activity (dpm/100 cm2)

Gross Beta Post-Remediation Surface Activity (dpm/100 cm2)

Total Removable 1

~12,000 2,400 1

2

~27,000 1,500 21 DM2 BMRC Confirmatory Survey Report C-1 5177-SR-01-0

Table C-2. Judgmental Measurement Locations for the Gamma Deck in the Reactor Containment Building Building Deck Areaa Reactor Building Containment Gamma Room 101A, Room 107, and Room109 Direct Measurement Number Pre-Remediation Gross Beta Surface Activity (dpm/100 cm2)

Gross Beta Post-Remediation Surface Activity (dpm/100 cm2)

Total Removable 3

~55,000

--a 4

~70,000

-220 1

5

~12,000 560 2

aContamination was removed by cutting out a portion of the wall.

BMRC Confirmatory Survey Report C-2 5177-SR-01-0

Table C-3. Judgmental Measurement Locations Associated With the Hot Cell Building Deck Areaa Reactor Building Containment Gamma Hot Cell and door Direct Measurement Number Pre-Remediation Gross Beta Surface Activity (dpm/100 cm2)

Gross Beta Post-Remediation Surface Activity (dpm/100 cm2)

Total Removable 6

~7,700

-520 1

7a

~6,300 6,200 3

8

~6,300 180 0

aArea was marked and will be removed as rad waste.

DM6 DM7 DM8 Approximately 6 kcpm; was marked and will be disposed of as rad waste Approximately 10 kcpm; was marked and will be disposed of as rad waste DM6 DM8 DM7 BMRC Confirmatory Survey Report C-3 5177-SR-01-0

Table C-4. Judgmental Measurement Locations for the Gamma Deck in the Administration Building Building Deck Area Administration Gamma Room 115A Direct Measurement Number Pre-Remediation Gross Beta Surface Activity (dpm/100 cm2)

Gross Beta Post-Remediation Surface Activity (dpm/100 cm2)

Total Removable 9

~3,800 3,600 6

BMRC Confirmatory Survey Report C-4 5177-SR-01-0

Table C-5. Judgmental Measurement Locations for the Control Deck in the Reactor Containment Building Building Deck Areaa Reactor Building Containment Control Room 202 Direct Measurement Number Pre-Remediation Gross Beta Surface Activity (dpm/100 cm2)

Gross Beta Post-Remediation Surface Activity (dpm/100 cm2)

Total Removable 10

~10,000 220a or 510b 1

11

~15,000 2,400 a or 5,500b 1

aTotal efficiency for Sr-90 used for surface activity calculation bTotal efficiency for Co-60 used for surface activity calculation DM10 DM11 Instrument response during scanning was approximately 1.7 kcpm Instrument response during scanning was approximately 2 kcpm BMRC Confirmatory Survey Report C-5 5177-SR-01-0

BMRC Radiological Survey Report C-6 5177-SR-01-0 Table C-6. Radionuclide Concentrations in Soil (pCi/g)

Sample ID Detector Response (cpm)

Ag-108m Co-60 Cs-137 Eu-152 Eu-154 Sr-90 SOF Pre-Sample Post-Sample 5177S0001 30,500 47,000 13.51 +/- 0.35a 3.13

+/- 0.14 0.61 +/- 0.06

-0.19

+/- 0.22

-0.03

+/- 0.04 7.87

+/- 0.57 7.1 5177S0002 25,000 33,000 10.13 +/-

0.62 2.60

+/- 0.19 0.61 +/- 0.08 0.04

+/- 0.16 0.02

+/- 0.07 5.51

+/- 0.50 5.2 5177S0003 25,000 32,000 7.66

+/-

0.46 1.75

+/- 0.14 0.39 +/- 0.05

-0.04

+/- 0.11 0.02

+/- 0.04 3.76

+/- 0.40 3.6 5177S0004 9,474 10,128 0.02

+/-

0.01 0.03

+/- 0.02 0.11 +/- 0.02

-0.03

+/- 0.06

-0.01

+/- 0.04 0.09

+/- 0.21 0.1 5177S0005 8,650 10,846 0.00

+/-

0.02 0.00

+/- 0.02 0.26 +/- 0.04 0.01

+/- 0.07

-0.02

+/- 0.04

-0.03

+/- 0.18 0.0 5177S0006 7,833 9,778 0.01

+/-

0.01

-0.01

+/- 0.03 0.07 +/- 0.02

-0.01

+/- 0.05

-0.01

+/- 0.02

--b 0.0 5177S0007 7,564 10,369 0.00

+/-

0.01 0.01

+/- 0.02 0.15 +/- 0.03

-0.04

+/- 0.05

-0.01

+/- 0.02

--b 0.0 5177S0008 7,819 10,518 0.00

+/-

0.02

-0.02

+/- 0.03 0.15 +/- 0.02 0.01

+/- 0.06 0.01

+/- 0.04 0.09

+/- 0.18 0.1 5177S0009 8,355 10,587 0.00

+/-

0.02

-0.03

+/- 0.04 0.15 +/- 0.03 0.02

+/- 0.07

-0.01

+/- 0.04

--b 0.0 5177S0010 8,509 10,425 0.00

+/-

0.02

-0.01

+/- 0.03 0.12 +/- 0.02

-0.02

+/- 0.06

-0.02

+/- 0.04

--b 0.0 aErrors represent the 95% uncertainty based on the total propagated uncertainty bNot analyzed; only S0001 to S0003 were requested for Sr-90 analysis by NRC, samples S0004, S0005, and S0008 were analyzed as reference values

APPENDIX D MAJOR INSTRUMENTATION BMRC Radiological Survey Report 5177-SR-01-0

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

D.1 ORAU 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, TX) coupled to:

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

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

D.1.2 BETA Ludlum Gas Proportional Detector Model 43-68, 126 cm2 physical area coupled to:

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

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

Ludlum Gas Proportional Detector Model 43-37, 582 cm2 physical area coupled to:

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

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

BMRC Radiological Survey Report D-1 5177-SR-01-0

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

Used in conjunction with:

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

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

Used in conjunction with:

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

Multichannel Analyzer Canberras Gamma Software Dell Workstation (Canberra, Meriden, CT)

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

BMRC Radiological Survey Report D-2 5177-SR-01-0

APPENDIX E SURVEY AND ANALYTICAL PROCEDURES BMRC Radiological Survey Report 5177-SR-01-0

E.1 PROJECT HEALTH AND SAFETY 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. All survey activities performed by ORAU were conducted in accordance with ORAU health and safety and radiation protection procedures (ORAU 2012; ORAU/ORISE 2011).

Pre-survey activities included the evaluation and identification of potential health and safety issues.

Survey work was performed per the ORAU generic health and safety plans and a site-specific Integrated Safety Management (ISM) pre-job hazard checklist.

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

Analytical and field survey activities were conducted in accordance with procedures from the following ORAU and ORAU/ORISE documents:

Survey Procedures Manual (ORAU/ORISE 2014a)

Laboratory Procedures Manual (ORAU/ORISE 2014b)

Quality Program Manual (ORAU 2014a)

The procedures contained in these manuals were developed to meet the requirements of 10 CFR 830 Subpart A, Quality Assurance Requirements and Department of Energy Order 414.1D Quality Assurance (CFR 2012 and DOE 2011).

Quality control procedures included:

Daily instrument background and check-source measurements to confirm that equipment operation was within acceptable statistical fluctuations Participation in Mixed-Analyte Performance Evaluation Program, NIST Radiochemistry Intercomparison Testing Program, and Intercomparison Testing Program Laboratory Quality Assurance Programs BMRC Radiological Survey Report E-1 5177-SR-01-0

Training and certification of all individuals performing procedures Periodic internal and external audits E.3 SURVEY PROCEDURES E.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 detectors (NaI) were not determined as the instruments were used solely as a qualitative means to identify elevated gamma radiation levels in excess of background. Identifications of elevated radiation levels that could exceed the site criteria were determined based on an increase in the audible signal from the indicating instrument.

Beta scans were performed using large area (floor monitor) gas proportional and small, hand-held gas proportional detectors with a 0.8 mg/cm-2 window. Identification of elevated radiation levels was based on increases in the audible signal from the indicating instrument. Beta surface scan MDCs were estimated using the approach described in NUREG-1507 (NRC 1998b). The scan MDC is a function of many variables, including the background level. Additional parameters selected for the calculation of scan MDCs included a two-second observation interval, a specified level of performance at the first scanning stage of 90% true positive and 25% false positive rate, which yields a d value of 1.96 (NUREG-1507, Table 6.1), and a surveyor efficiency of 0.5. The beta total weighted efficiency based on Tc-99 for Co-60 was 0.11. The average concrete background for the detectors was around 327 counts per minute (cpm). The minimum detectable count rate (MDCR) and scan MDC was calculated as:

Bi = (327)(2 s)(1 min/60 s) = 11 counts MDCR = (1.96)(11 counts)1/2[(60 s/min)/2s] = 195 cpm MDCRsurveyor = 195/(0.5)1/2 = 276 cpm Scan MDC = (276)/(0.11*1.26) = 1,991 dpm/100 cm 2 (~2,000 dpm/100 cm2)

E.3.2 SURFACE ACTIVITY MEASUREMENTS Measurements of gross beta surface activity levels were performed using hand-held gas proportional detectors coupled to portable ratemeter-scalers. Count rates (cpm), which were integrated over one BMRC Radiological Survey Report E-2 5177-SR-01-0

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. The gross beta efficiency was 0.11 (calibrated with Tc-99). ORAU determined construction material-specific background for each surface type encountered for determining net count rates. The a priori MDC for beta activity is given by:

= 3 + 4.65 Where:

B

=

background tot

=

total efficiency G

=

geometry correction factor (1.26)

The a priori static MDC for concrete at the BMRC was 630 dpm/100 cm2.

E.3.3 REMOVABLE ACTIVITY MEASUREMENT Removable gross beta activity levels were determined using numbered filter paper disks, 47 mm in diameter. Moderate pressure was applied to the smear and approximately 100 cm2 of the surface was wiped. Smears were placed in labeled envelops with the location and other pertinent information recorded.

E.3.3 SOIL SAMPLING Soil sampling was performed by NRC staff. Approximately 0.5 to 1 kg of soil was collected at each sample location. Collected samples were placed in a plastic bag, sealed, and labeled in accordance with ORAU/ORISE survey procedures. The judgmental soil samples were collected as individual samples from areas of elevated gamma radiation based on gamma scans.

E.4 RADIOLOGICAL ANALYSIS E.4.1 GAMMA SPECTROSCOPY Samples of soil were dried, 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 and volumes were determined and the samples counted using intrinsic germanium detectors coupled to a pulse height BMRC Radiological Survey Report E-3 5177-SR-01-0

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) that were associated with the radionuclides of concern were reviewed for consistency of activity. TAPs used for determining the activities of the radionuclides of concern and the typical associated MDCs for a four-hour count time were as follows.

Radionuclide TAPa (MeV)

MDC (pCi/g)

Co-60 1.332 0.06 Cs-137 0.662 0.05 Ag-108m 0.434 0.04 Eu-152 0.344 0.20 Eu-154 0.123 0.09 aSpectra were also reviewed for other identifiable total absorption peaks (TAPs) that would not be expected at this site.

MDC = minimum detectable concentration.

E.4.1 Sr-90 ANALYSIS Soil samples were dissolved by a combination of potassium hydrogen fluoride and pyrosulfate fusions. The fusion cake was dissolved and strontium was coprecipitated 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.

E.4.2 GROSS ALPHA/GROSS BETA ANALYSES Smears were counted on a low-background gas proportional system for gross alpha and beta activity.

The minimum detectable activities of the procedure were 11dpm and 14 dpm for alpha and beta activity respectively.

BMRC Radiological Survey Report E-4 5177-SR-01-0

E.5 UNCERTAINTIES The uncertainties associated with the analytical data presented in the tables of this report represent the total propagated uncertainties for that data. These uncertainties were calculated based on both the gross sample count levels and the associated background count levels.

E.6 DETECTION LIMITS Detection limits, referred to as MDCs, were based on 95% confidence level via 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 Radiological Survey Report E-5 5177-SR-01-0